[Federal Register Volume 64, Number 8 (Wednesday, January 13, 1999)]
[Notices]
[Pages 2243-2259]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-660]
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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 December 18, 1998, through December 31,
1998. The last biweekly notice was published on December 30, 1998 (63
FR 71962).
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 February 12, 1999, 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
[[Page 2244]]
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner who fails
to file such a supplement which satisfies these requirements with
respect to at least one contention will not be permitted to participate
as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
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 amendments request: November 30, 1998.
Description of amendments request: Currently, the Calvert Cliffs
Technical Specifications allow defective tubes to be plugged and
removed from service, or to be repaired by either the laser-welded
sleeving technique developed by Westinghouse Electric Corporation or by
using leak-tight, tungsten inert gas-welded sleeving developed by
Combustion Engineering, Inc. (ABB-CE). The proposed amendment will
revise the appropriate Technical Specifications to permit the use of
leak-limiting Alloy 800 repair sleeves developed by ABB-CE to be used
at Calvert Cliffs. Combustion Engineering provides two types of leak-
limiting Alloy 800 repair sleeves. The first type of repair sleeve
spans the expansion transition zone of the tube at the top of the
tubesheet. The second type of repair sleeve spans the degraded areas at
an eggcrate support elevation or in a free span section.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment would not involve a significant increase
in the probability or consequences of an accident previously evaluated.
The ABB CE Alloy 800 leak-limiting repair sleeves are designed
using the applicable American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code and, therefore, meet the design
objectives of the original steam generator tubing. The applied stresses
and fatigue usage for the repair sleeves are bounded by the limits
established in the ASME Code. Mechanical testing has shown that the
structural strength of repair sleeves under normal, upset, and faulted
conditions provides margin to the acceptance limits. These acceptance
limits bound the most limiting (three times normal operating pressure
differential) burst margin recommended by Regulatory Guide 1.121. Burst
testing of sleeved tubes has demonstrated that no unacceptable levels
of primary-to-secondary leakage are expected during any plant
condition.
The Alloy 800 repair sleeve Technical Specification depth-based
plugging limit is determined using the guidance of Regulatory Guide
1.121 and the pressure stress equation of ASME Code, Section III. A
bounding tube wall degradation growth rate per cycle and a
nondestructive examination uncertainty has been assumed for determining
the repair sleeve plugging limit.
Evaluation of the repaired steam generator tubes indicates no
detrimental effects on the sleeve or sleeve-tube assembly from reactor
system flow, primary or secondary coolant chemistries, thermal
conditions or transients, or pressure conditions as may be experienced
at Calvert Cliffs. Corrosion testing of sleeve-tube assemblies
indicates no evidence of sleeve or tube corrosion considered
detrimental under anticipated service conditions.
The implementation of the proposed amendment has no significant
effect on either the configuration of the plant, or the manner in which
it is operated. The consequences of a hypothetical failure of the
sleeved tube is bounded by the current steam generator tube rupture
analysis described in Calvert Cliffs Updated Final Safety Analysis
Report, Section 14.15. Due to the slight reduction in diameter caused
by the sleeve wall thickness, primary coolant release rates would be
slightly less than assumed for the steam generator tube rupture
analysis and, therefore, would result in lower total primary fluid mass
release to the secondary system. A main steam line break or feed line
break will not cause a SGTR [steam generator tube rupture] since the
sleeves are analyzed for a maximum accident differential pressure
greater than that predicted in the Calvert Cliffs safety analysis. The
minimal repair sleeve leakage that could occur during plant operation
is well within the Technical Specification leakage limits.
Therefore, BGE has concluded that the proposed change does not
involve a significant increase in the probability or consequences of an
accident previously evaluated.
2. Would not create the possibility of a new or different kind of
accident from any other accident previously evaluated.
As discussed above, the Alloy 800 repair sleeves are designed using
the applicable ASME Code as guidance; therefore, it meets the
objectives of the original steam generator tubing. As a result, the
functions of the steam generators will not be significantly affected by
the installation of the proposed sleeve. The proposed repair sleeves do
not interact with any other plant systems. Any accident as a result
[[Page 2245]]
of potential tube or sleeve degradation in the repaired portion of the
tube is bounded by the existing tube rupture accident analysis. The
continued integrity of the installed sleeve is periodically verified by
the Technical Specification requirements.
The implementation of the proposed amendment has no significant
effect on either the configuration of the plant, or the manner in which
it is operated. Therefore, BGE [Baltimore Gas and Electric Company]
concludes that this proposed change does not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Would not involve a significant reduction in a margin of safety.
The repair of degraded steam generator tubes with Alloy 80 leak-
limiting repair sleeves restores the structural integrity of the
degraded tube under normal operating and postulated-accident
conditions. The design safety factors utilized for the repair sleeves
are consistent with the safety factors in the ASME Boiler and Pressure
Vessel Code used in the original steam generator design. The portions
of the installed sleeve assembly that represent the reactor coolant
pressure boundary can be monitored for the initiation and progression
of sleeve/tube wall degradation. Use of the previously identified
design criteria and design verification testing assures that the margin
to safety is not significantly different from the original steam
generator tubes.
Therefore, BGE concludes that the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: S. Singh Bajwa, Director.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: November 9, 1998.
Description of amendment request: The proposed amendments would
revise Technical Specification Table 3.3.3-2, ``Emergency Core Cooling
System Actuation Instrumentation Setpoints'' to modify the degraded
voltage second level undervoltage relay setpoint and allowable value.
This change was submitted in response to a concern identified during an
Electrical Distribution System Functional Inspection.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The setpoint change does not change the logic or function of the
degraded voltage protection circuits as described in UFSAR [Updated
Final Safety Analysis Report] Section 8.2.3. They also do not reduce
the reliability of these circuits. The increase in the degraded voltage
protection circuit setpoint is conservative compared to the existing
setpoint. There is no change as a result of this amendment to the
underlying accident and transient analyses that support operations of
LaSalle County Station. Inadvertent or spurious operation of the
degraded voltage protection function will initiate loading of the safe
shutdown loads on the diesel generators and is not assumed to initiate
an accident. The proposed degraded voltage setpoints are low enough to
prevent spurious actuations given the expected offsite grid voltages.
After implementation of this amendment, no operator actions are
required for equipment operations in response to degraded voltage
conditions.
This change does not affect the initiators or precursors of any
accident previously evaluated. This change will not increase the
likelihood that a transient initiating event will occur because
transients are initiated by equipment malfunction and/or catastrophic
system failure.
The consequences of accidents previously evaluated are not
increased. The proposed change does not affect the required level of
availability of systems required to mitigate the accidents considered
in the analyses. The proposed changes will ensure that the Class 1E
equipment will be capable of starting and operating during a design
basis accident with degraded offsite grid voltage. The increase in the
level of confidence is the result of more rigorous methodology used to
determine limiting Class 1E bus voltages at the minimum expected
offsite AC voltage. These calculations demonstrate that the degraded
voltage relays will not actuate following a block start of the
electrical loads that are automatically actuated by or as a consequence
of the LOCA [loss-of-coolant accident] signal if the switchyard voltage
remains above 352 kV.
If the grid voltage drops below 352 kV, then the analytical limit
of 3814 volts for proper operation of class 1E loads connected to each
4.16 kV Class 1E bus is assured by transfer to the respective onsite
power sources (Emergency Diesel Generators (EDGs)) by the degraded
voltage logic.
Therefore this proposed amendment does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
(2) Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
Setpoint methodology established the bases to ensure that, with
known errors, the relays will detect degraded voltage conditions and
transfer safety loads to the EDGs at a voltage level adequate to ensure
proper safety equipment performance and to prevent equipment damage.
The greater than or equal to 3870 volt setpoint and the greater
than or equal to 3814 volt allowable value includes adequate tolerance
to calibrate the relay trip units while ensuring that the Class 1E bus
voltage will remain above the analytical limits.
These setpoint changes will ensure that adequate voltages will be
available for the continuous operation of safety-related equipment
required to function during a LOCA. These proposed changes will also
ensure that adequate voltages will be available for starting any Class
1E equipment.
The proposed degraded voltage setpoint change does not change the
design of the degraded voltage protection system or its function to
protect against degraded offsite power. Actuation of the degraded
voltage protection system will initiate a sequence of events that will
start the EDG for the associated Class 1E bus, strip loads from the
Class 1E bus, open all feed breakers to the Class 1E bus, close the
Emergency feed breaker (thus energizing the Class 1E bus from the
respective EDG), and initiate starting of the Safe Shutdown equipment
supplied by the Class 1E bus.
Since the scope of this change does not affect the operation of the
auxiliary power system or any actions necessary to mitigate the
consequences of accidents or achieve safe shutdown, the
[[Page 2246]]
change does not involve a new or different accident scenario.
Therefore, these proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
(3) Involve a significant reduction in the margin of safety
because:
The proposed amendment will allow the degraded voltage setpoint to
be conservatively established based on new engineering calculations
which consider the lowest expected offsite grid voltage and operation
of required Class 1E equipment under design basis accident loading
conditions.
The proposed degraded voltage setpoints will ensure that adequate
Class 1E bus voltage will be available to support starting and
operation of the required Class 1E loads. The proposed setpoint
includes instrument error to ensure that the lowest possible voltage
will not be lower than the degraded voltage analytical limits.
Additionally, the proposed setpoints are low enough to prevent spurious
actuations due to expected fluctuations in the grid voltage. The new
setpoints are also set with margin to the minimum Class 1E bus voltage,
which is based on a minimum grid voltage of 352 kV, which is less than
the expected grid voltage of 354 kV. The proposed changes will provide
an increase in the level of protection that currently exists and will
ensure the margin of safety is adequately maintained.
Therefore, these changes do not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library, 815
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,
Illinois 61348-9692.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Stuart A. Richards.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois
Date of amendment request: November 30, 1998.
Description of amendment request: The Safe Shutdown Makeup Pump
(SSMP) allowed outage time (AOT) is being decreased from 67 days to 14
days.
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 change to the Technical Specification Allowed Outage Time is
conservative with respect to current requirements. This change is being
proposed to establish an AOT for the SSMP that is equivalent to that
for the reactor core isolation cooling (RCIC) pump (14 day AOT) in
order to enhance system performance by assuring maximum SSMP pump
availability to a level consistent with RCIC. This is necessary since,
pursuant to Paragraph III.G.3 of 10 CFR 50, Appendix R, the SSMP is an
alternate system to the RCIC system. By ensuring equipment
availability, the probability or consequences of an accident previously
evaluated are not increased. In addition, the proposed change has no
impact on any accident initiators or 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.
Therefore, this proposed amendment does not involve a significant
increase 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 a reduction to a Technical
Specification Allowed Outage Time to enhance system performance by
assuring maximum SSMP pump availability to a level consistent with
RCIC. The proposed change is conservative with respect to the 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 change 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. The proposed change enhances system performance by
assuring maximum SSMP pump availability to a level consistent with
RCIC. Since this is a conservative change that will enhance the
performance of the SSMP system, it does not involve a significant
reduction in the margin of safety.
Therefore, this 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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: 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.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3 (CR-3), Citrus County, Florida
Date of amendment request: October 30, 1998.
Description of amendment request: The proposed amendment would
change the Crystal River Unit 3 (CR-3) Improved Technical
Specifications (ITS) to delete a note regarding the number of required
channels for the Degrees of Subcooling function, and to subdivide the
Core Exit Temperature (Backup) function into two new functions in ITS
Table 3.3.17-1, Post-Accident Monitoring Instrumentation.
These proposed ITS changes support modifications scheduled for
Refueling Outage 11 at CR-3. These modifications are intended to
significantly improve the reliability and availability of information
to the control room operators for verifying adequate core cooling is
maintained following a design basis accident. The proposed ITS change
deletes the note describing the use of the SPDS as a backup since the
SPDS will be the primary indication of subcooling margin after the
planned modifications are implemented.
[[Page 2247]]
The planned modifications will separate the sixteen core exit
thermocouples into two separate channels of eight core exit
thermocouples each. Following the modifications, there will be two core
exit thermocouples per channel located in each core quadrant. Each
separate channel of eight core exit thermocouples will have an
associated core exit temperature recorder on the main control board,
instead of the current three recorders, and will provide input into the
associated channel of SPDS for calculation of subcooling margin.
The proposed ITS change will subdivide the current Core Exit
Temperature (Backup) function into two new functions, Core Exit
Temperature (Thermocouple) function and Core Exit Temperature
(Recorder) function. For the Core Exit Temperature (Thermocouple)
function, the proposed ITS will require at least two OPERABLE core exit
thermocouples per core quadrant (at least one per channel) to provide a
representative distribution of temperatures across the core to the
operator. For the Core Exit Temperature (Recorder) function, both core
exit temperature recorders will be required OPERABLE.
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 changes to the Degrees of Subcooling, Core Exit
Temperature (Thermocouple), and Core Exit Temperature (Recorder)
functions in the CR-3 Improved Technical Specifications (ITS) ensure
appropriate post-accident monitoring instrumentation is available for
use by the operators during implementation of emergency operating
procedures. These emergency operating procedures provide direction to
the operators for ensuring that actions required to mitigate the
effects of the previously evaluated design basis accidents are
performed. The instrumentation is used for monitoring by the operators
after an accident occurs, perform no automatic functions, and there are
no credible failures of this instrumentation which could initiate any
accident previously evaluated. Therefore, the probability of occurrence
of any accident previously evaluated is unaffected.
The availability and use of this instrumentation ensures that the
prescribed manual operator actions for mitigating the consequences of
an accident will be implemented when necessary, and that the operator
has sufficient information to verify required automatic actions have
occurred when necessary. The availability and use of the
instrumentation provides assurance that the consequences of accidents
will not be greater than that previously evaluated. The associated
modifications that are planned for these post-accident monitoring
instruments will enhance the reliability of the required indications to
the operators.
2. Create the possibility of a new or different kind of accident
from previously evaluated accidents?
The proposed changes to this post-accident monitoring
instrumentation will ensure appropriate instrumentation is available
for use by the operators following a design basis accident. This
instrumentation is necessary for performing certain manual actions, or
to verify automatic actions have occurred, which are required to
mitigate the effects of a design basis accident. The instrumentation is
used for monitoring by the operators after an accident occurs, perform
no automatic functions, and there are no credible failures of this
instrumentation which could initiate a new or different kind of
accident. Therefore, the possibility of a new or different kind of
accident occurring as a result of this passive instrumentation is not
created.
3. Involve a significant reduction in a margin of safety?
The proposed changes to this post-accident monitoring
instrumentation provide additional assurance that adequate
instrumentation is available for use by the operators to perform manual
actions, and to verify that automatic actions that are required to
mitigate the effects of a design basis accident have occurred. The
instrumentation is used for monitoring by the operators after an
accident occurs, and perform no automatic functions. The availability
and use of this instrumentation ensures that the prescribed manual
operator actions for mitigating the consequences of an accident will be
implemented when necessary, and that the operators have sufficient
information to verify required automatic actions have occurred when
necessary. These required manual and automatic actions are necessary to
preserve the margin of safety as defined in the CR-3 ITS. The
availability and use of this instrumentation provides assurance that
the existing margin of safety will be maintained, and assumptions
related to the margin of safety during mitigation of design basis
accidents will be preserved. Therefore, the existing margin of safety
will not be reduced.
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: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC--A5A, P.O. Box 14042, St. Petersburg, Florida
33733-4042.
NRC Project Director: Frederick J. Hebdon.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3 (CR-3), Citrus County, Florida
Date of amendment request: November 24, 1998.
Description of amendment request: The proposed amendment would
change the CR-3 Improved Technical Specifications (ITS) in support of a
modification to install a diesel-driven emergency feedwater (EFW) pump
(EFP-3) which is intended to resolve capacity limitations of the CR-3
1A Emergency Diesel Generator (EGDG). The licensee has determined that
installation of EFP-3 involves an unreviewed safety question and also
requires changes and additions to the ITS and Bases.
EFP-3 will be installed as a functional replacement for EFP-1, the
motor-driven EFW pump. EFP-3 will start and provide controlled and
monitored EFW flow to both steam generators through the same EFW block
and control valves as EFP-1 currently uses. The licensee stated that
removing the auto-start logic from EFP-1 would eliminate the need to
perform EGDG-1A load management to accommodate emergency safeguards
(ES) loads required to mitigate design basis accidents. EFP-1 will
remain available as a manually started pump. The installation of EFP-3
will also permit other changes in system operation which are intended
to reduce reliance on operator actions to perform EGDG load management.
The proposed ITS and Bases changes fall into two categories: (1)
new or revised ITS and Bases to account for equipment changes
associated with the new EFP-3, and (2) those ITS and Bases requirements
being deleted because they were approved until Cycle 12 only.
[[Page 2248]]
The new ITS requirements and revisions (category 1 changes) involve
revised surveillance requirements (SR) and Bases for EFP-3 (SR 3.7.5),
and new ITS and Bases for the diesel fuel oil supply, lube oil and
starting air for EFP-3 (3.7.19). Also, the option to use EFP-3 for Once
Through Steam Generator (OTSG) cooling is added to the Bases for 3.4.6,
RCS Loops--MODE 5, Loops Filled. The Bases for 3.4.6, Background, lists
all feedwater pumps that may be available in MODE 5. EFP-3 is added
here for completeness.
The Bases for 3.7.5 are revised to describe the new EFP-3 and the
new role for EFP-1 as a manual defense-in-depth pump. The Bases are
also revised to indicate that EFP-3 cannot directly access the
condenser hotwell. The phrase ``with the exception of the loss of all
AC power (Ref. 3)'' is deleted from the Applicable Safety Analysis
because with the addition of EFP-3, the EFW system is able to maintain
its function on a loss of off-site power (LOOP) with a single failure.
In Section 3.7.5, EFW System, one SR is being revised and one new
SR is being added. These changes are intended to provide SRs that
demonstrate OPERABILITY of EFP-3 and essential subsystems. SR 3.7.5.1
and Bases are revised to add verification of proper valve position for
starting air and fuel oil flow paths for EFP-3 on a 45-day frequency.
SR 3.7.5.6 is added to provide assurance that the DC electrical
support system will be available to support OPERABILITY of EFP-3. This
SR is based on a similar SR currently approved for the station DC
system required by ITS 3.8.4, DC Sources--Operating. SR 3.7.5.6 was
determined necessary because DC power is essential for starting EFP-3.
ITS 3.7.19 was added to ensure essential subsystems are within
limits needed to maintain EFP-3 OPERABLE. The specification includes
requirements for fuel oil, lube oil, and starting air. This
specification has an allowed outage time (AOT) for these parameters if
they are less than the limit but above a minimum value. Below the
minimum allowed value, EFP-3 must be declared inoperable.
A number of ITS and Bases are being revised to remove the
requirements that permitted operation of CR-3 until Cycle 12 only
(category 2 changes). All text marked with the footnote ``Note--Valid
until Cycle 12 only,'' and the note itself, is being deleted except for
a few instances which are discussed in the licensee's submittal.
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.
4. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
This change involves the addition of a new safety-related Diesel-
Driven Emergency Feedwater Pump (EFP-3). The Emergency Feedwater (EFW)
System is not an initiator for any design basis accident except for
those accidents associated with an increase in primary to secondary
cooling and a loss of heat sink. The new EFP-3 functionally replaces
the Motor-Driven Emergency Feedwater Pump (EFP-1) and is no more likely
to cause an inadvertent cooldown than the existing EFP-1. The starting
logic of the Emergency Feedwater Initiation and Control system is the
same for EFP-3 as it was for EFP-1 before. No other control or logic
changes are being made that would make EFP-3 more likely to cause a
cooldown transient.
EFP-3 has a slightly greater probability of failing to start
compared to EFP-1 with offsite power available. Therefore, there is a
slight increase in the probability of an event that involves a loss of
heat sink when considering only the Improved Technical Specifications
(ITS) required EFW Pumps. The new EFP-3 will be highly reliable and
therefore this increase in risk is not significant. Loss of EFP-3 alone
does not cause a total loss of heat sink without the loss of the
Turbine-Driven Emergency Feedwater Pump (EFP-2) and the remaining
feedwater pumps. The most important of these feedwater pumps is EFP-1,
which will be maintained as a safety-grade backup. EFP-3 is less
reliable than EFP-1 with offsite power available. However, if offsite
power is not lost, EFP-1 should be available for use. Therefore, the
overall EFW system reliability is enhanced.
The consequences of the failure of EFP-3 to start or inadvertently
actuate were considered. Failure of EFP-3 to start will have the same
impact as failure of EFP-1. Therefore, the consequences of evaluated
accidents are the same. EFP-3 will be designed to have minimum and
maximum flows equivalent to EFP-1. No changes to the system will cause
a decrease in the ability of the EFW system to remove heat from the
Once Through Steam Generators (OTSGs). Similarly, the heat removal
capability of EFP-3 will not be different than EFP-1. Therefore, there
will not be the potential of a significantly greater overcooling event
due to inadvertent start of EFP-3.
The license changes associated with the addition of EFP-3 remove a
number of ITS Actions that established compensatory measures due to the
possibility of overloading the Emergency Diesel Generators (EGDGs) and
cross-train dependencies with EFP-2. These compensatory actions are no
longer required. The changes to the EFW system eliminate EGDG
limitations and reliance of the ``A'' train EFW pump on EFP-2. The
revised ITS Actions ensure the equipment required to mitigate an
accident is restored to OPERABLE status in accordance with previously
approved limits. In addition, replacing required operator actions with
automatic functions provides greater assurance that mitigating actions
will occur. Therefore, these changes will not adversely affect the
probability or consequences of evaluated accidents.
Based on the above, the addition of EFP-3 and the associated
license changes do not involve a significant increase in the
probability or consequences of a previously evaluated accident.
5. Does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
EFP-3 performs the same functions as the existing EFP-1. No plant
conditions are changed to cause new or different accidents. Although a
diesel engine has different failure modes than a motor-driven pump, the
consequences of a pump failure are the same. An interlock and
administrative controls are provided to ensure that both EFP-3 and EFP-
1 do not run at the same time. The interlock and administrative
controls prevent any new interactive failure modes that could be caused
by having both ``A'' train pumps (or all three EFW pumps) operating at
the same time.
The revised ITS Actions ensure equipment is restored to OPERABLE
status in accordance with previously approved timeframes. No new plant
configurations or conditions are created by these Actions.
Therefore, these changes cannot create the possibility of an
accident of a different type than previously evaluated in the SAR
[Safety Analysis Report].
6. Does not involve a significant reduction in the margin of
safety.
EFP-3 is designed to meet the same performance criteria as EFP-1.
EFP-3 will replace EFP-1 in the ITS. The pump will perform the same
functions, will be reliable and meet the same design criteria. There
are no functions performed by EFP-1 that will be significantly
different with EFP-3. The
[[Page 2249]]
margin of safety provided by the specification relates to the ability
to provide a heat sink. EFP-3 will provide the same margin of safety.
In addition, EFP-1 will be available as a safety-grade backup and can
deliver EFW to the OTSGs if offsite power is available or if the ``A''
train EGDG has adequate load margin.
The cooling capability of EFP-3 will be equivalent to EFP-1.
Therefore, EFP-3 provides the same protection to the fuel cladding from
temperature excursions as EFP-1. The EFP-3 modifications will be done
without making penetrations through reactor coolant system (RCS) or
containment boundaries. Therefore, the integrity of these fission
product barriers remains unchanged.
The proposed changes to the ITS delete temporary restrictions
placed on systems due to the potential to overload the EGDGs and cross-
train dependencies with EFP-2. These compensatory actions are no longer
required. The changes to the EFW system eliminate EGDG limitations and
reliance of the ``A'' train EFW pump on EFP-2. The revised ITS Actions
ensure the equipment required to mitigate an accident is restored to
OPERABLE status in accordance with previously approved limits. In
addition, replacing required operator actions with automatic functions
provides greater assurance that mitigating actions will occur.
Based on the above evaluation, there is no reduction in the margin
of safety associated with the proposed equipment, system and license
changes.
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: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC--A5A, P. O. Box 14042, St. Petersburg, Florida
33733-4042.
NRC Project Director: Frederick J. Hebdon.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of amendment request: July 14, 1998.
Description of amendment request: The proposed amendment would
amend the Technical Specifications to revise the liquid and gaseous
release rates to reflect the replacement of the former 10 CFR 20.106
requirements with the existing 10 CFR 20.1302 requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change does not:
1. Involve a significant increase in the probability or consequence
of an accident previously evaluated.
The likelihood that an accident will occur is neither increased nor
decreased by these Technical Specification changes. These Technical
Specifications changes will not impact the function or method of
operation of plant equipment. No systems, equipment, or components are
affected by the proposed changes. The proposed revisions to the liquid
and gaseous release rate limits will not result in any change or
increase in the types or amounts of effluents other than that which has
historically been deemed acceptable for release, nor will there be an
increase in individual or cumulative occupational radiation exposures
other than that which has historically been deemed acceptable.
Therefore, the proposed changes to the Technical Specifications do not
involve any increase in the probability or consequences of any 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 involve changes to the physical plant
or operations. The proposed changes are administrative in nature and
will not change the types and amounts of effluents from that which has
historically been deemed acceptable. Since these administrative changes
do not contribute to accident initiation, they do not produce a new
accident scenario nor do they alter any existing accident scenarios.
Therefore, the proposed changes to the Technical Specifications would
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 because
compliance with the limits of the existing 10 CFR 20.1301 will be
demonstrated by operating within the limits of 10 CFR Part 50, Appendix
I and 40 CFR Part 190. For the liquid effluent releases the annual dose
of 500 mrem, upon which the concentrations in the previous 10 CFR Part
20, Appendix B, Table II, Column 2, are based, is a factor of 10 higher
than the annual dose of 50 mrem, upon which the concentrations in the
existing 10 CFR 20, Appendix B, Table 2, Column 2, are based. Also, for
gaseous effluent releases, the limits associated with the gaseous
release Technical Specifications will be revised to the previously
acceptable instantaneous dose rate limits.
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: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578.
Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic
Power Company, 321 Ferry Road, Wiscasset, ME 04578.
NRC Project Director: Seymour H. Weiss.
Northeast Nuclear Energy Company, (NNECO) et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of amendment request: December 10, 1998.
Description of amendment request: The proposed amendment would
allow NNECO to implement plant modifications that would ensure that
proper flow paths can be established for boron precipitation control
after a loss-of-coolant accident.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
In accordance with 10 CFR 50.92, NNECO [Northeast Nuclear Energy
Company] has reviewed the proposed changes and has concluded that they
do not involve a significant hazards consideration (SHC). The basis for
this conclusion is that the three criteria of 10 CFR 50.92(c) are not
compromised. The proposed changes do not involve an SHC because the
changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed plant modifications will ensure proper flow paths can
be established for boron precipitation control after a Loss of Coolant
Accident
[[Page 2250]]
(LOCA). This will be accomplished by the following plant modifications:
a. Provide an alternate AC source of power for 2-SI-651, ``Shutdown
Cooling Header Containment Isolation Valve,'' a Facility Z1 component,
from Facility Z2.
b. Provide an alternate DC source of power for 2-CH-517,
``Auxiliary Spray Charging Header Supply Valve,'' a Facility Z2
component, from Facility Z1.
c. Provide an alternate DC source of power for 2-CH-519, ``Loop 1A
Charging Header Supply Valve,'' a Facility Z2 component, from Facility
Z1.
d. Provide test jacks to determine valve position for LPSI [low-
pressure safety injection] injection valves 2-SI-615, 2-SI-625, 2-SI-
635, and 2-SI-645 at the respective motor control center (MCC [motor
control center] B51 for 2-SI-615 and 2-SI-625 (MCC B61 for 2-SI-635 and
2-SI-645).
e. Provide bypass capability of the low pressure open permissive
for 2-SI-651.
The alternate power supply to valves 2-SI-651, 2-CH-517 and 2-CH-
519, and the position indication for valves 2-SI-615, 2-SI-625, 2-SI-
635, and 2-SI-645 cannot initiate an accident. The proposed
modifications will not change the design parameters, failure positions
or design requirements of the valves. The proposed plant modifications
will ensure valves 2-SI-651, 2-CH-517 and 2-CH-519 can operate after a
LOCA to perform their accident mitigating functions. Therefore,
providing an additional power source to 2-SI-651, 2-CH-517 and 2-CH-
519, and a local means of determining the position of valves 2-SI-615,
2-SI-625, 2-SI-635, and 2-SI-645 cannot initiate an accident and will
not adversely affect the function of these components to mitigate the
consequences of an accident.
The proposed plant modifications will also bypass the open
permissive for 2-SI-651. This pressure permissive, which protects the
low pressure Shutdown Cooling (SDC) System from the high pressure
Reactor Coolant System (RCS), allows 2-SI-651 to be opened only when
pressurizer pressure is below 280 psia [pounds per square inch
absolute]. This pressure permissive would be disabled upon a loss of
Facility Z1 power. This would prevent the opening of 2-SI-651. The new
local control switch for 2-SI-651, which bypasses this pressurizer
pressure permissive, is isolated by normally open relay contacts. When
aligned to its alternate power, local control is enabled, and remote
control in the Main Control Room is isolated. Multiple operator errors
would be required to align 2-SI-651 to the alternate power source
during normal operation. To misalign these valves, an equipment
operator would have to perform steps located only in an Emergency
Operating Procedure. Additionally, control room operators would have to
disregard annunciators that indicate the valves are being transferred
to their alternate power source. Therefore, the only time the valve is
expected to be opened by the local control switch is after a LOCA with
a Facility Z1 failure. Currently, the potential exists for an operator
to open the valve when pressure is above 280 psia. An undetectable
single failure of the contact which provides the permissive would allow
an operator to open the valve even when pressure is above 280 psia.
During normal operation, this condition would be annunciated in the
Main Control Room. During accident conditions, this annunciator may be
disabled. Therefore, 2-SI-651 could be opened with pressurizer pressure
above 280 psia without annunciation. Although 2-SI-651 could be opened,
the pressure permissive for 2-SI-652, the upstream isolation valve
(Attachment 1 Figure 1), would prevent 2-SI-652 from opening. This
would protect the shutdown cooling suction line from
overpressurization. During accident conditions, if both valves were
opened and pressure increased above 280 psia, annunciation of 2-SI-652
being open would be available to provide indication of the potential
overpressure condition. Therefore, the installation of the capability
to bypass the open permissive for 2-SI-651 will not result in a
significant increase in the probability or consequences of an accident
previously evaluated.
The proposed plant modifications have no adverse effect on how any
of the associated systems or components function to prevent or mitigate
the consequences of design basis accidents. Also, the proposed changes
have no adverse effect on any design basis accident previously
evaluated since the modifications will ensure that accident mitigation
equipment will be available to function as assumed in the LOCA
analysis. Therefore, the proposed plant modifications do not result in
a significant increase in the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed plant modifications will provide the capability of
powering 2-SI-651, 2-CH-517 and 2-CH-519 from either Facility Z1 or
Facility Z2, and will add test jacks to MCC B51 and B61 to determine
the position of valves 2-SI-615, 2-SI-625, 2-SI-635, and 2-SI-645.
Additionally, this activity adds a local control switch which will
bypass the open permissive for 2-SI-651 when aligned to the alternate
power source. A single failure in any of the breakers or disconnect
switches which allow 2-SI-651, 2-CH-517 and 2-CH-519 to be powered from
either facility is bounded by the failure of the valve. A failure of
any of the test jacks may result in a loss of control power to the
associated valves. This failure is also bounded by the failure of the
valve. During normal operation the local control switch which bypasses
the pressure permissive is isolated by normally open contacts. A single
failure of the local control switch or isolating relay during normal
operation cannot disable the pressure permissive.
Since a single failure of any component added by this activity is
bounded by existing component failures, a failure of these components
cannot create a new accident. Therefore, the proposed plant
modifications will 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 plant modifications will ensure boron precipitation
control can be established. This will be accomplished by providing an
alternate power source for 2-SI-651, 2-CH-517 and 2-CH-519, and adding
test jacks to determine the position of valves 2-SI-615, 2-SI-625, 2-
SI-635, and 2-SI-645. Additionally, this activity adds a local control
switch which will bypass the open permissive for 2-SI-651 when aligned
to the alternate power source. Although the potential exists to route
redundant power trains in the same cable trays, conduits and cable, the
design of the modifications ensures that a single failure will not
compromise the redundant power distribution system. The installation of
the connection jacks and local control switch will not alter the
failure analysis for the valves, and will not change the design
parameters of the valves (i.e. pressure rating). Therefore, the
proposed plant modifications will not compromise RCS pressure
boundaries, containment integrity, or fuel cladding. In addition, the
new disconnect switches, breakers, cabling, and auxiliary components
are all designed for the rated voltages and currents, and are QA
[quality assurance] Category I seismically and environmentally
qualified, as required.
[[Page 2251]]
Based on the above, the proposed plant modifications will not
reduce the integrity of the plant protective boundaries, or adversely
affect the LOCA analysis. These modifications will have no adverse
effect on equipment important to safety. The equipment will continue to
function as assumed in the design basis accident analysis. This will
ensure that the acceptance criteria of 10 CFR 50.46(b)(5) for long term
core cooling will be met. Therefore, there will be 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 Project Director: William M. Dean.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of amendment request: December 16, 1998.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Surveillance Requirements 4.8.1.1.2
and 4.8.1.1.3, Table 4.8.1.1.2-1, and the associated Bases. The
proposed changes would remove the Emergency Diesel Generator
accelerated testing and special reporting requirements from the TSs in
accordance with the guidance provided in Generic Letter 94-01.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed TS changes do not involve any physical changes to
plant structures, systems, or components. [Public Service Electric &
Gas Company] PSE&G has implemented the provisions of the Maintenance
Rule for diesel generators, including the associated regulatory
guidance, thereby establishing a program that assures diesel generator
performance. The elements of the program include the performance of
detailed root cause analysis of individual failures, effective
corrective actions taken in response to individual failures, and
implementation of preventive maintenance consistent with the
Maintenance Rule. Monitoring the effectiveness of diesel generator
maintenance and continuing surveillance testing in accordance with the
proposed TS changes will ensure that the diesel generators will perform
their intended functions and will minimize failures. The accelerated
testing requirements are therefore no longer considered to be necessary
and are deleted. The requirements of 10 CFR 50.72 and 10 CFR 50.73
ensure that diesel generator failures are properly reported. The
special reporting requirements are therefore unnecessary and are
deleted. Based on the above information, the changes will not adversely
affect the assurance of diesel generator reliability or operability,
and there is no significant increase in the probability or consequences
of any accident previously evaluated.
(2) The proposed change does not create the possibility of a new or
different kind of accident from any accident previously analyzed.
The proposed TS changes do not involve any physical changes to the
design of plant systems, structures or components, nor do the changes
involve a change in plant operation. The diesel generators will
continue to function as designed to mitigate the consequences of an
accident. Eliminating the accelerated testing requirements and special
reporting requirements does not permit plant operation in a
configuration that would create a different type of malfunction to the
diesel generators than any previously evaluated. In addition, the
proposed TS changes do not alter the conclusions described in the
[Updated Final Safety Analysis Report] UFSAR regarding the safety
related functions of the diesel generators or their support systems. No
new failure modes will be introduced. Therefore, the proposed 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.
This request does not involve an adverse impact on diesel generator
design, operation, or reliability. Since monitoring and maintenance is
being performed in conformance with 10 CFR 50.65, modifying the
surveillance testing frequency requirements does not adversely affect
the reliability of the diesel generators. Deletion of the special
reporting requirements does not impact operability or reliability of
the diesel generator. Since the diesel generator function is not
affected by the proposed change, this request 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: 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.
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.
Indiana Michigan Power Company, Docket, Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: December 3, 1998.
Brief description of amendments: The amendments would revise
Technical Specification Section 4.6.5.1, ``Ice Condenser, Ice Bed,''
and the associated bases to reflect the maximum ice condenser flow
channel blockage assumed in the accident analyses.
[[Page 2252]]
Date of publication of individual notice in Federal Register:
December 28, 1998.
Expiration date of individual notice: January 27, 1999.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085.
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.
Arizona Public Service Company, et al., Docket Nos. STN 50-528 and STN
50-529, Palo Verde Nuclear Generating Station, Units Nos. 1, 2, and 3,
Maricopa County, Arizona
Date of application for amendment: October 6, 1998.
Brief description of amendment: The amendments revise Technical
Specifications (TS) 3.3.1, ``Reactor Protective System (RPS)
Instrumentation--Operating,'' and TS 3.3.2, ``Reactor Protective System
(RPS) Instrumentation--Shutdown.'' The amendments clarify the power
level threshold at which certain RPS instrumentation trips must be
enabled and may be bypassed, and clarify that this level is a
percentage of the neutron flux at rated thermal power (RTP). The bypass
power level, 1E-4% RTP, is specified as logarithmic power instead of
thermal power. The NRC approved these changes for Palo Verde Unit 3 on
an exigent basis in its letter dated October 19, 1998. The exigent TS
amendment resulted in TS pages with notes specifying different
requirements between Unit 3 and Units 1 and 2. These amendments remove
these notes regarding Unit 3 from the affected TS pages so that all
Units now have the same TS.
Date of issuance: December 23, 1998.
Effective date: December 23, 1998.
Amendment No.: Unit 1-119; Unit 2-119.
Facility Operating License Nos. NPF-41 and NPF-51: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: November 4, 1998 (63 FR
59586).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 23, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of application for amendment: February 11, 1998.
Brief description of amendment: The Updated Final Analysis Report
(UFSAR) describes the response of the salt service water (SSW) system
to a complete loss of AC power by assuming that the system would be
divided by the closure of one of the two division isolation valves.
Boston Edison Company (BECo) has discovered single failures involving a
partial loss of AC power could place the SSW system in a configuration
of one pump supplying both trains of heat exchangers for the first 10
minutes of the worst case design basis accident. BECo has determined
that these single failures are an unreviewed safety question. The
amendment authorizes BECo to change UFSAR Section 10.7, ``Salt Service
Water System,'' to address this single fauilure vulnerability.
Date of issuance: December 21, 1998.
Effective date: December 21, 1998.
Amendment No.: 180.
Facility Operating License No. DPR-35: Amendment revised the UFSAR.
Date of initial notice in Federal Register: April 8, 1998, (63 FR
17220)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 21, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: March 10, 1997, as supplemented
May 23, 1997, and October 15, 1998.
Brief description of amendment: This amendment revises Technical
Specification (TS) 3.5.1, ``Emergency Core Cooling System (ECCS)
Accumulators,'' by (1) increasing the allowed outage time (from 1 hour
to 72 hours) that one ECCS accumulator can be inoperable as a result of
the boron concentration being outside of TS limits, and (2) modifying
surveillance requirement 4.5.1 consistent with the guidance provided in
NUREG-1366, ``Improvements to Technical Specifications Surveillance
Requirements,'' December 1992, and the Standard Technical
Specifications (STS) for Westinghouse Plants, NUREG-1431, Revision 1.
Date of issuance: December 31, 1998.
Effective date: December 31, 1998.
Amendment No.: 86.
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications
Date of initial notice in Federal Register: April 9, 1997 (62 FR
17226).
The May 23, 1997, and October 15, 1998, submittals contained
clarifying information only, and did not change the initial no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 31, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
[[Page 2253]]
Consumers Energy Company, Docket No. 50--155, Big Rock Point (BRP)
Plant, Charlevoix County, Michigan
Date of application for amendment: September 19, 1997.
Brief Description of amendment: This amendment changes the DPR--6
License and revises its Technical Specifications to reflect the
permanently shutdown and defueled condition of the BRP plant.
Date of issuance: December 24, 1998.
Effective date: No later than 45 days from date of issuance.
Amendment No.: 120.
Facility Operating License No. DPR-6: The amendment revises the
DPR--6 License and Appendix A Technical Specifications to the licensee.
Date of initial notice in Federal Register: December 3, 1997 (62 FR
63974).
No significant hazards consideration comments received: No.
Local Public Document Room location: North Central Michigan College
Library, 1515 Howard Street, Petoskey, MI 49770.
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: July 13, 1998.
Brief description of amendments: These amendments change the Beaver
Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and BVPS-2) Updated
Final Safety Analysis Reports (UFSAR) descriptions of the Intake
Structure main entrance and interconnecting cubicle doors. The changes
approved by these amendments address a new failure mode of safety-
related equipment that had not been previously considered for BVPS-1.
The changes state that the cubicle interconnecting flood protection
doors are normally closed with their inflatable seals depressurized and
that the associated security/fire doors are normally closed. This door
closure arrangement provides protection for the safety-related
equipment in the interconnecting cubicles from the consequences of
potential internal flooding.
Date of issuance: December 16, 1998.
Effective date: December 16, 1998.
Amendment Nos.: 218 and 96.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
approve changes to the Updated Final Safety Analysis Reports.
Date of initial notice in Federal Register: August 12, 1998 (63 FR
43202)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 16, 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 No. 50-334, Beaver Valley Power
Station, Unit No. 1, Shippingport, Pennsylvania
Date of application for amendment: June 18, 1996, as supplemented
September 8 and 30, 1998.
Brief description of amendment: The amendment (1) makes editorial
changes to Technical Specification (TS) 4.4.5 and associated Bases; (2)
revises the Bases for TS 3.4.6.2 to provide consistency with the Beaver
Valley Power Station, Unit No. 1, Updated Final Safety Analysis Report
(UFSAR); and (3) revises Index Page XVII to reflect the revision of
page numbers due to shifting of text by License Amendment No. 198.
Date of issuance: December 21, 1998.
Effective date: As of date of issuance, to be implemented within 60
days.
Amendment No: 219.
Facility Operating License No. DPR-66. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 64109).
The September 8 and 30, 1998, letters did not change the initial
proposed no significant hazards consideration determination or expand
the amendment request beyond the scope of the November 18, 1998,
Federal Register notice; these letters only provided updated TS pages
to be consistent with the UFSAR.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 21, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: May 31, 1996.
Brief description of amendment: The amendment revises the
surveillance test interval for the reactor protection system reactor
trip breakers, reactor trip modules, and electronic trip relays from a
monthly interval to a quarterly interval.
Date of issuance: December 31, 1998.
Effective date: December 31, 1998.
Amendment No.: 194.
Facility Operating License No. DPR-51: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 28, 1996 (61 FR
44356).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 31, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
2, Pope County, Arkansas
Date of amendment request: May 18, 1998, as supplemented on
December 8, 1998.
Brief description of amendment: The amendment allows the use of
trisodium phosphate stored in three baskets on the containment floor as
a replacement to the sodium hydroxide addition system for the control
of sump pH during long term core cooling in recirculation phase.
Date of issuance: December 23, 1998.
Effective date: The license amendment is effective as of its date
of issuance to be implemented prior to the facility's restart from
refueling outage 2R13.
Amendment No.: 194.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56241)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 23, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: July 28, 1997.
Brief description of amendment: The amendment modifies the actions
associated with Technical Specification (TS) Table 3.3-1 for the
Reactor Protective Instrumentation and TS Table 3.3-3 for the
Engineered Safety Feature Actuation System Instrumentation.
Date of issuance: December 29, 1998.
Effective date: December 29, 1998, to be implemented within 30
days.
[[Page 2254]]
Amendment No.: 195.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 27, 1997 (62 FR
45456).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 29, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: June 29, 1998, as supplemented
by letters dated December 17, 1998 and December 22, 1998.
Brief description of amendment: This amendment revises the as-found
lift setting tolerance for the ANO-2 main steam safety valves and the
pressurizer safety valves, revises the maximum allowable linear power
level-high trip setpoint with inoperable steam line safety valves, and
relocates part of the specifications for steam line safety valves to
the ANO-2 Safety Analysis Report. Administrative and bases changes have
also been made.
Date of issuance: December 31, 1998.
Effective date: The license amendment is effective as of its date
of issuance to be implemented within 30 days.
Amendment No.: 197.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56242).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 31, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: August 12, 1998.
Brief description of amendment: The amendment changes the Technical
Specifications by increasing the maximum boron concentration in the
Safety Injection Tanks (SITs) and the Refueling Water Storage Pool
(RWSP) from 2300 ppm to 2900 ppm.
Date of issuance: December 21, 1998.
Effective date: December 21, 1998, to be implemented within 60
days.
Amendment No.: 147.
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56249).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 21, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: June 11, 1996, and supplemented
March 26, 1997.
Brief description of amendments: The amendments relocate certain
quality assurance related requirements from the TS to the licensee's
Quality Assurance Program Description.
Date of issuance: December 28, 1998.
Effective date: December 28, 1998, with full implementation within
120 days.
Amendment Nos.: 226 and 210.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 31, 1996 (61 FR
40022).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 28, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085.
Indiana Michigan Power Company, Docket No. 50-315 , Donald C. Cook
Nuclear Plant, Unit 1, Berrien County, Michigan
Date of application for amendment: August 28, 1998, as supplemented
November 4, 1998.
Brief description of amendment: The amendment grants relief from
the steam generator surveillance requirement in Section 4.4.5.3 of the
Technical Specifications (TS). The surveillance requirement is
associated with non-destructive examination of the steam generator
tubes which is required every 24 months. The relief allows the
examination to be deferred from April 8, 1999, until the next refueling
outage for D.C. Cook, Unit 1.
Date of issuance: December 30, 1998.
Effective date: December 30, 1998, with full implementation within
45 days.
Amendment No.: 227.
Facility Operating License No. DPR-58: Amendment adds paragraph
2.C.(9) to the License.
Date of initial notice in Federal Register: October 7, 1998 (63 FR
53950).
The November 4, 1998, submittal provided additional information
that did not change the initial no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 30, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: April 30, 1997, as supplemented
November 12, 1998.
Brief description of amendment: The amendment deletes TSs
requirements associated with meterological monitoring instrumentation
which have been relocated to the Updated Safety Analysis Report in
accordance with 10 CFR 50.36 and the guidance in NRC Generic Letter 95-
10, ``Relocation of Selected Technical Specification Requirements
Related to Instrumentation.''
Date of issuance: December 22, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 85.
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: June 18, 1997 (62 FR
33126).
The November 12, 1998, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 22, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
[[Page 2255]]
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of application for amendment: September 19, 1995.
Brief description of amendment: The amendment reduces the frequency
of the Technical Specification (TS) 4.5.1.d surveillance interval for
boron concentration of the safety injection tasks from once per 31 days
to once every 6 months. Initially, the change was requested for TS
Section 4.5.1.b. However, TS Section 4.5.1.b was subsequently changed
to TS Section 4.5.1.d by Amendment No. 220 to Facility Operating
License No. DPR-65 dated September 3, 1998, in response to NNECO's
application dated August 23, 1995.
Date of issuance: December 17, 1998.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 221.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 25, 1995 (60 FR
54722).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 17, 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. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: July 2, 1998.
Brief description of amendment: The amendment revises the Updated
Final Safety Analysis Report (UFSAR) by changing UFSAR Sections 9.7.2,
``Service Water,'' and 9.4, Reactor Building Closed Cooling Water,'' to
include in the discussions the use of various types of internal
protective coatings and liners used in the piping and components of the
systems. The change also indicates that periodic maintenance,
surveillance, and inspections will be conducted to ensure that coating
or liner degradation will be promptly detected and corrected to provide
reasonable assurance that the systems can perform their safety-related
functions.
Date of issuance: December 18, 1998.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 222.
Facility Operating License No. DPR-65: Amendment revised the
Updated Final Safety Analysis Report and Appendix B to Operating
License.
Date of initial notice in Federal Register: August 12, 1998 (63 FR
43206).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 18, 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.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: July 5, 1995, as supplemented
October 9, 1998.
Brief description of amendment: The amendment extends surveillance
test intervals and allowable out-of-service times for instrumentation
in the Emergency Core Cooling (ECCS), Rod Block, Isolation Group 4
(High Pressure Coolant Injection, or HPCI) and Isolation Group 5
(Reactor Core Isolation Cooling, or RCIC), Reactor Building Ventilation
& Standby Gas Treatment, Recirculation Pump Trip and Alternate Rod
Injection, and Shutdown Cooling Supply Isolation Systems.
Date of issuance: December 23, 1998.
Effective date: December 23, 1998, with full implementation within
30 days.
Amendment No.: 103.
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45182). The October 9, 1998, submittal withdrew a portion of the
original request, made additional editorial changes, and provided
updated Technical Specification pages. 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 December 23, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: August 15, 1996, as supplemented
March 19 and October 12, 1998.
Brief description of amendment: The amendment revises the Technical
Specifications so that either 8 or 12 hour shifts will be considered
``normal'' and 40 hours will be considered a ``nominal'' week, changes
the wording for surveillances required ``once per shift'' to ``once per
12 hours,'' clarifies the ``once per hour'' wording related to fire
watch patrols, and makes a number of other clarifications and
typographical corrections.
Date of issuance: December 24, 1998.
Effective date: December 24, 1998, with full implementation within
30 days.
Amendment No.: 104.
Facility Operating License No. DPR-22: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 7, 1998 (63 FR
53951). The October 12, 1998, submittal provided additional
clarifications and new TS pages. 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 December 24, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
[[Page 2256]]
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: December 14, 1995, as
supplemented on November 25, 1996, April 10, September 4, and December
29, 1997, January 8, March 2, June 11, August 12, and October 30, 1998.
Brief description of amendments: The amendments revise Technical
Specifications (TS) Table of Contents; TS 3.1, ``Reactor Coolant
System;'' TS 4.0, ``Surveillance Requirements;'' TS 5.0, ``Design
Features;'' and associated Bases by removing or relocating requirements
that are adequately controlled by existing regulations other than 10
CFR 50.36 and the TS and by modifying TS 6.0 to more closely meet the
format and content of the standard technical specifications.
Date of issuance: December 7, 1998.
Effective date: December 7, 1998, with full implementation of the
TS and License Condition 7 by September 1, 1999. License Condition 6
shall be implemented by the next USAR update, but no later than June 1,
1999. Implementation shall also include the relocation of TS
requirements to the appropriate licensee-controlled documents as
identified in the licensee's application dated December 14, 1995, as
supplemented on November 25, 1996, April 10, September 4, and December
29, 1997, January 8, March 2, June 11, August 12, and October 30, 1998,
and evaluated in the staff's safety evaluation attached to these
amendments.
Amendment Nos.: 141 and 132.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Licenses and TS.
Date of initial notice in Federal Register: June 5, 1996 (61 FR
28618). The November 25, 1996, April 10, September 4, and December 29,
1997, January 8, March 2, June 11, August 12, and October 30, 1998,
submittals provided additional clarifying information, revised
implementation dates, and updated TS pages. 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 amendments is contained
in a Safety Evaluation dated December 7, 1998. `
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: September 4, 1998.
Brief description of amendments: The amendments revise Technical
Specifications (TS) 3.1.A.3.b, 4.18, and Bases for TS 4.18 to clarify
the surveillance requirements and limiting conditions for operation of
the reactor coolant vent system.
Date of issuance: December 17, 1998.
Effective date: December 17, 1998, with full implementation within
30 days.
Amendment Nos.: 142 and 133.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the TS.
Date of initial notice in Federal Register: September 23, 1998 (63
FR 50938).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 17, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: April 17, 1997.
Brief description of amendment: The amendment revises Technical
Specifications (TS) 2.12, ``Control Room Systems,'' to delete the
limiting condition for operation (LCO) and surveillance for control
room temperature and replace it with an associated LCO and surveillance
for the control room air conditioning system. In addition, the
amendment revises TS 2.1, ``Reactor Coolant System,'' TS 2.6,
``Containment System,'' and TS 2.8, ``Refueling Operations,'' and the
associated surveillance requirements to incorporate the design basis
requirements for refueling operations and to correspond to NUREG-1432,
``Standard Technical Specifications Combustion Engineering Plants.''
Date of issuance: December 31, 1998.
Effective date: December 31, 1998, to be implemented within 60 days
from the date of issuance.
Amendment No.: 188.
Facility Operating License No. DPR-40. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 4, 1997 (62 FR
30639).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 31, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
Power Authority of the State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: April 16, 1998, as supplemented
August 20, 1998.
Brief description of amendment: The amendment will extend the
surveillance interval for five instrument channels from the current 18
months to 24 months. The proposed amendment also revises Section 6 of
the Technical Specifications to reflect updated analyses.
Date of issuance: December 16, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 185.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56256).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 16, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: July 6, 1998.
Brief description of amendment: The amendment revises Appendix B
Technical Specification 3.5, Main Condenser Steam Jet Air Ejector and
Table 3.10-1, Radiation Monitoring Systems that Initiate and /or
Isolate Systems including the associated Bases to provide Allowable
Outage Times for selected instrumentation.
[[Page 2257]]
Date of issuance: December 28, 1998.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 249.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 12, 1998 (63 FR
43211).
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated December 28, 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.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of application for amendments: August 1, 1997, as supplemented
on October 6, 1997, February 18 and July 7, 1998.
Brief description of amendments: The amendments revise Technical
Specification Section 4.2.1 of Appendix B to require that Public
Service Electric & Gas Company (PSE&G) adhere to the Incidental Take
Statement, approved by the National Marine Fisheries Service (NMFS),
but remove the specific requirements. Removing the specific
requirements of Section 4.2.1 enables PSE&G to utilize relief granted
by the NMFS on a case-by-case basis.
Date of issuance: December 18, 1998.
Effective date: Effective as of its date of issuance, to be
implemented within 60 days.
Amendment Nos: 216 and 196.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 10, 1997 (62
FR 47698).
The October 6, 1997, February 18 and July 7, 1998 submittals
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 December 18, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of application for amendment: May 21, 1997, as supplemented on
December 4, 1998. The December 4, 1998, submittal contained clarifying
information only, and did not change the initial no significant hazards
consideration determination.
Brief description of amendment: The amendment revises the Virgil C.
Summer Nuclear Station Technical Specifications to change the methods
for testing the control room and spent fuel pool ventilation system
charcoal adsorbers from American National Standards Institute Standard
N509-1980 to American Society for Testing and Materials Standard D3803-
1989.
Date of issuance: December 23, 1998.
Effective date: December 23, 1998.
Amendment No.: 140.
Facility Operating License No. NPF-12: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: June 18, 1997 (62 FR
33133).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 23, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San
Diego County, California
Date of application for amendments: May 7, 1998.
Brief description of amendments: The amendments revise the
reference for obtaining the thyroid dose conversion factors used in the
definition of Dose Equivalent Iodine 131 (I-131) in Technical
Specification Section 1.1, ``Definitions.''
Date of issuance: December 16, 1998.
Effective date: December 16, 1998, to be implemented within 30 days
from the date of issuance.
Amendment Nos.: Unit 2--145; Unit 3--137.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 4, 1998 (63 FR
59595).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 16, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San
Diego County, California
Date of application for amendments: June 12, 1998, as supplemented
by letters dated September 18, 1998, October 29, 1998, and November 23,
1998.
Brief description of amendments: The amendments authorize revision
of the San Onofre Nuclear Generating Station Updated Final Safety
Analysis Report to incorporate a new turbine missile protection
calculation methodology.
Date of issuance: December 21, 1998.
Effective date: December 21, 1998, to be implemented in the next
periodic update of the UFSAR in accordance with 10 CFR 50.71(e) that
occurs after 60 days of the date of issuance.
Amendment Nos.: Unit 2-146; Unit 3-138.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
authorize revisions to the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: November 9, 1998 (63 FR
60412).
The November 23, 1998, supplemental letter provided additional
information and did not change the original no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 21, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear
Plant, Unit No. 1, Docket No. 50-260, Browns Ferry Nuclear Plant, Unit
No. 2, and Docket No. 50-296, Browns Ferry Nuclear Plant, Unit No. 3,
Limestone County, Alabama
Date of amendment request: June 2, 1997 as supplemented November
19, 1998.
Description of amendment request: Presently Technical Specification
(TSs) require both the recirculation loops to
[[Page 2258]]
be operable and provide a 12-hour allowable outage time (AOT) for
single loop operation (SLO) mode. The amendments modify TS to allow
indefinite SLO instead of the 12-hour AOT.
Date of issuance: December 23, 1998.
Effective date: December 23, 1998.
Amendment No.: 236, 256 and 216.
Facility Operating License Nos. DPR-33, DPR-52, and DPR-68.
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 13, 1997 (62 FR
43377). The licensee's letter of November 19, 1998, did not expand the
scope of the application or affect the staff's initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 23, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: February 18, 1998.
Brief description of amendment: Changes Technical Specification
(TS) 3.7.4, Steam Generator Atmospheric Dump Valves (ADVs), and its
associated bases by adding a new TS CONDITION, REQUIRED ACTION, and
COMPLETION TIME to address a potential condition where two ADVs are
made technically inoperable when one train of the safety-related
auxiliary control air system is taken out of service.
Date of issuance: December 17, 1998.
Effective date: December 17, 1998.
Amendment No.: 16.
Facility Operating License No. NPF-90: Amendment revises the TSs.
Date of initial notice in Federal Register: August 12, 1998 (63 FR
43213) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 17, 1998.
No significant hazards consideration comments received: None.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: February 28, 1996, as
supplemented October 2 and December 12, 1997, March 30 and December 11,
1998.
Brief description of amendment: The February 28, 1996 letter
proposed to extend the surveillance interval for Westinghouse type AR
relays with alternating current and direct current coils from quarterly
to an 18 month interval. The letter of December 11, 1998 revised the
scope of the application such that it now applies only to Westinghouse
type AR relays which use alternating current coils. Accordingly, this
amendment approves the extension of the surveillance interval only for
Westinghouse type AR relays which use alternating current coils.
Date of issuance: December 30, 1998.
Effective date: December 30, 1998.
Amendment No.: 17.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 10, 1996 (61 FR
15998). The October 2 and December 12, 1997, March 30 and December 11,
1998 letters provided clarifying information that did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 30, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402.
The Cleveland Electric Illuminating Company, Centerior Service Company,
Duquesne Light Company, Ohio Edison Company, OES Nuclear, Inc.,
Pennsylvania Power Company, Toledo Edison Company, Docket No. 50-440
Perry Nuclear Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: June 30, 1998, as supplemented
by submittals dated October 27, November 30, and December 3, 1998. The
supplemental submittals did not expand the scope of the original
application or change the staff's proposed no significant hazards
considerations determination.
Brief description of amendment: This amendment reflects the
approval of the transfer of the authority to operate the Perry Nuclear
Power Plant, Unit 1, under the license to a new company, FirstEnergy
Nuclear Operating Company. In addition, several administrative changes
unrelated to the transfer are being made to delete certain sections of
the license relating solely to one-time historical events that have
occurred.
Date of issuance: December 21, 1998.
Effective date: December 21, 1998.
Amendment No.: 96.
Facility Operating License No. NPF-58: This amendment revised the
operating license.
Date of initial notice in Federal Register: August 4, 1998 (63 FR
41600).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 21, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: June 29, 1998, as supplemented
by submittals dated July 14, October 26, and November 30, 1998.
Brief description of amendment: This amendment reflects the
approval of the transfer of the authority to operate Davis-Besse
Nuclear Power Station, Unit 1, under the license to a new company,
FirstEnergy Nuclear Operating Company.
Date of issuance: December 21, 1998.
Effective date: December 21, 1998.
Amendment No.: 228.
Facility Operating License No. NPF-3: Amendment revised the
operating license.
Date of initial notice in Federal Register: August 4, 1998 (63 FR
41602) The additional information provided in the supplemental
submissions provided clarifying information only which did not affect
the staff's proposed no significant hazards consideration or expand the
scope of the application as noticed initially.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 21, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
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: August 2, 1996 (TXX-96434), as
supplemented by
[[Page 2259]]
letters dated October 2, 1998 (TXX-98215), and November 13, 1998 (TXX-
98241 and TXX-98244).
Brief description of amendments: The amendment increases the
allowed outage time (AOT) for a centrifugal charging pump from 72 hours
to 7 days and adds a Configuration Risk Management Program.
Date of issuance: December 29, 1998.
Effective date: December 29, 1998, to be implemented within 30
days.
Amendment Nos.: Unit 1--Amendment No. 62; Unit 2--Amendment No. 48.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 27, 1998, (63
FR 65617) supersedes FR notice dated September 24, 1997.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 29, 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.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: May 8, 1998, as supplemented on
July 10 and October 2, 1998.
Brief description of amendment: The amendment reduces the normal
operating suppression pool water temperature limit and adds a time
restriction for the temperature limit allowed during surveillances that
add heat to the suppression pool.
Date of Issuance: December 28, 1998.
Effective date: December 28, 1998, to be implemented within 30
days.
Amendment No.: 163.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 23, 1998 (63
FR 50941).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated December 28, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: September 12, 1996, as
supplemented April 24, 1997, and September 24, 1998
Brief Description of amendments: The amendments revise License
Condition 3.I, Fire Protection, and relocate fire protection
requirements from the Technical Specifications to the Updated Final
Safety Analysis Report.
Date of issuance: December 16, 1998.
Effective date: December 16, 1998.
Amendment Nos.: 217 and 217.
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
change the Licenses and Technical Specifications.
Date of initial notice in Federal Register: November 4, 1998 (63 FR
59598).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 16, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: October 10, 1996, as
supplemented by letter dated November 9, 1998.
Brief description of amendment: The amendment changes Facility
Operating License No. NPF-21 to authorize the storage of byproduct,
source, and special nuclear materials at the WNP-2 site. These
materials had been originally stored at the WNP-1 site and are not
intended for use at WNP-2.
Date of issuance: December 29, 1998.
Effective date: December 29, 1998, to be implemented within 45 days
from the date of issuance.
Amendment No.: 155.
Facility Operating License No. NPF-21: The amendment revised the
operating license.
Date of initial notice in Federal Register: September 23, 1998 (63
FR 50942).
The November 9, 1998, supplemental letter provided additional
clarifying information that did not change the staff's original no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 29, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Dated at Rockville, Maryland, this 6th day of January 1999.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 99-660 Filed 1-12-99; 8:45 am]
BILLING CODE 7590-01-P