[Federal Register Volume 63, Number 231 (Wednesday, December 2, 1998)]
[Notices]
[Pages 66590-66609]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-31931]


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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 November 6, 1998, through November 19, 1998. 
The last biweekly notice was published on November 18, 1998 (63 FR 
64106).

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.

[[Page 66591]]

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 January 4, 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 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.

Duke Energy Corporation (DEC), et al., Docket Nos. 50-413 and 50-414, 
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: November 11, 1998.

[[Page 66592]]

    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) to correct Surveillance 
Requirements (SRs) 3.6.11.6 and 3.6.11.7 and the associated Bases. 
These SRs currently are incorrect and do not reflect the Containment 
Pressure Control System (CPCS) as designed. Therefore, the proposed 
amendments would only revise the SRs; no change to the CPCS design is 
involved.
    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:

First Standard

    Implementation of this amendment would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. Approval of this amendment will have no 
significant effect on accident probabilities or consequences.
    The CPCS is not an accident initiating system; therefore, there 
will be no impact on any accident probabilities by the approval of 
this amendment. The design of the CPCS is not being modified by this 
proposed amendment. The amendment merely aligns [TS] surveillance 
requirements with the existing design and function of the system. 
Therefore, there will be no impact on any accident consequences.

Second Standard

    Implementation of this amendment would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. No new accident causal mechanisms are created 
as a result of NRC approval of this amendment request. No changes 
are being made to the plant which will introduce any new accident 
causal mechanisms. This amendment request does not impact any plant 
systems that are accident initiators, since the CPCS is an accident 
mitigating system.

Third Standard

    Implementation of this amendment would not involve a significant 
reduction in a margin of safety. Margin of safety is related to the 
confidence in the ability of the fission product barriers to perform 
their design functions during and following an accident situation. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment system. The performance of these fission 
product barriers will not be impacted by implementation of this 
proposed amendment. The CPCS is already capable of performing as 
designed. No safety margins will be impacted.

    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: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.
    Attorney for licensee: Mr. Paul R. Newton, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina.
    NRC Project Director: Herbert N. Berkow.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of amendment request: October 15, 1998.
    Description of amendment request: The proposed amendments would 
revise the pressure-temperature limits in the Technical Specifications 
for Units 1, 2, and 3. The proposed amendments would revise the heatup, 
cooldown, and inservice test limitations for the reactor coolant system 
of each unit to a maximum of 26 effective full-power years. The 
proposed amendments would also revise the Technical Specification for 
low temperature overpressure protection to reflect the revised 
pressure-temperature limits of the reactor vessels.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    NO.
    Each accident analysis addressed in the Oconee UFSAR [Updated 
Final Safety Analysis Report] has been examined with respect to the 
changes to the Reactor Pressure Vessel (RPV) pressure-temperature 
limit curves and related Low Temperature Overpressure settings. The 
probability of any design basis accident (DBA) is not affected by 
this change, nor are the consequences of a DBA affected by this 
change. The revised pressure-temperature limits, which were 
developed based on NRC approved methodology or ASME Code [American 
Society of Mechanical Engineers Boiler and Pressure Vessel Code] 
Case N-514 as described in the Technical Justification, are not 
considered to be an initiator or contributor to any accident 
analysis addressed in the Oconee UFSAR. The added requirement to 
deactivate one pressurizer heater bank during low temperature 
operation does not significantly change the probability or 
consequence of any accident previously analyzed. No existing 
Technical Specification requirements are being deleted with this 
revision.
    B. Create the possibility of a new or different kind of accident 
from the accident previously evaluated?
    NO.
    This license amendment revises Oconee RPV pressure-temperature 
limits. The revised pressure-temperature limits were developed based 
on NRC approved methodology or ASME Code Case N-514 as described in 
the Technical Justification. Operation of Oconee in accordance with 
these proposed new Technial Specifications will not create any 
failure modes not bounded by previously evaluated accidents. 
Consequently, this change will not create the possibility of a new 
or different accident from any accident previously evaluated.
    C. Involve a significant reduction in a margin of safety?
    NO.
    This license amendment revises Oconee RPV pressure-temperature 
limits. The revised pressure-temperature limits were developed based 
on NRC approved methodology or ASME Code Case N-514 as described in 
the Technical Justification. The purpose of this license amendment 
is to assure that sufficient operating margin to safety is 
maintained in the operation of the Oconee reactor pressure vessels 
by establishing new, more limiting pressure-temperature limit curves 
and adding the requirement to deactivate one pressurizer heater 
bank. No plant safety limits, set points, or design parameters are 
adversely affected. The fuel, fuel cladding, and Reactor Coolant 
System are not impacted. Therefore, there will be no significant 
reduction in any margin of safety.
    Duke [Duke Energy Corporation] has concluded based on this 
information that there are no significant hazards considerations 
involved in this amendment request.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC.
    NRC Project Director: Herbert N. Berkow. Duquesne Light Company, et 
al., Docket No. 50-334, Beaver Valley Power Station, Unit No. 1, 
Shippingport, Pennsylvania

    Date of amendment request: November 11, 1998.
    Description of amendment request: The proposed amendment would 
modify License Condition 2.C(9) to allow, on a one time only basis, an 
extension to the steam generator inspection interval of technical 
specification surveillance 4.4.5.3.b. This

[[Page 66593]]

would allow the steam generator inspection interval to coincide with 
the 13th refueling outage or the end of 500 effective full power days, 
whichever occurs sooner.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change is temporary and allows a one time extension 
of specific surveillance requirements for Cycle 13 to allow 
surveillance testing to coincide with the 13th (1R13) refueling 
outage. The proposed surveillance interval extension will not cause 
a significant reduction in system reliability nor affect the ability 
of a system to perform its design function. Current monitoring of 
plant conditions and the surveillance monitoring required during 
normal plant operation will be performed as usual to assure 
conformance with technical specification operability requirements.
    The technical specification steam generator tube inspection is 
intended to prevent the Steam Generator Tube Rupture analyzed in 
[Updated Final Safety Analysis Report] UFSAR Section 14.2.4 by 
maintenance of the integrity of the primary to secondary coolant 
boundary represented by steam generator tubes. The process by which 
this integrity is maintained is inspection of steam generator tubes 
at prescribed intervals, and the removal of defective tubes from 
service. Inspection intervals are based on preventing corrosion 
growth from exceeding tube structural limits, thereby preventing 
tube failure. The 1997 steam generator inspection characterized 
existing steam generator tube degradation, and degraded tubes were 
removed from service at that time. Degradation growth rates were 
evaluated for the next operating interval and it was determined that 
the steam generator tube structural integrity is maintained. 
Degradation of steam generator tubes was prevented during the 
extended outage by a carefully controlled, corrosion prevention 
program.
    The proposed change does not affect the UFSAR and is consistent 
with changes granted for other plants. The surveillance extension 
does not involve a change to plant equipment and does not affect the 
performance of plant equipment used to mitigate an accident. This 
change, therefore, does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Extending the surveillance interval for the performance of 
specific inspections will not create the possibility of any new or 
different kind of accidents. No change is required to any system 
configurations, plant equipment or analyses.
    Steam generator tube inspections determine tube integrity and 
provide reasonable assurance that a tube rupture or primary to 
secondary leak will not occur. Accidents involving steam generator 
tube rupture are analyzed in UFSAR Section 14.2.4, ``Steam Generator 
Tube Rupture.'' The only type of accident that can be postulated 
from extending the steam generator inspection interval would be a 
tube leak or rupture which are analyzed in the UFSAR. No new failure 
modes are created by the surveillance extension. Therefore, this 
change will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Surveillance interval extensions will not impact any plant 
safety analyses since the assumptions used will remain unchanged. 
The safety limits assumed in the accident analyses and the design 
function of the equipment required to mitigate the consequences of 
any postulated accidents will not be changed since only the 
surveillance interval is being extended. Based on engineering 
judgement, extending the surveillance interval for the performance 
of these specific inspections does not involve a significant 
reduction in the margin of safety derived from the required 
surveillances.
    The margin of safety depends upon maintenance of specific 
operating parameters within design limits. In the case of steam 
generators, that margin is maintained through assurance of tube 
integrity as the primary to secondary boundary. Assurance of tube 
integrity is provided through periodic in-service inspection of 
tubes and removal of defective tubes from service. Additional margin 
is provided through protection from possible consequences of steam 
generator tube failure by mitigation systems. Radiation monitors 
provide a detection capability of primary to secondary leakage to 
enable a prompt response. Maintenance of the steam generator water 
chemistry in accordance with [Electric Power Research Institute] 
EPRI guidelines provides additional margin of safety. Therefore, the 
plant will be maintained within the analyzed limits and the proposed 
extension will not significantly reduce the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.
    Attorney for Licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Robert A. Capra Entergy Operations, Inc., 
Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County, 
Arkansas
    Date of amendment request: June 30, 1998.
    Description of amendment request: The proposed change modifies the 
Engineered Safety Features Actuation System (ESFAS) portion of the 
Arkansas Nuclear One, Unit-2 (ANO-2) Plant Protection System (PPS). 
This modification is designed to defeat the backup power supply for the 
auctioneered power sources for channel A and D Reactor Protective 
System (RPS) and ESFAS bistables, and to provide selective logic for 
Emergency Feedwater Actuation Signals and Main Steam Isolation Signals. 
This will ensure that ESFAS will have the redundancy and independence 
sufficient to assure that (1) no single failure results in loss of the 
protection function with a channel in indefinite bypass, and (2) 
removal from service of any component or channel does not result in 
loss of the required minimum redundancy required by the ANO-2 Technical 
Specifications (TSs). The proposed modification to the ANO-2 PPS has 
been determined to involve an Unreviewed Safety Question in accordance 
with 10 CFR 50.59(a)(2).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    An evaluation of the proposed change has been performed in 
accordance with 10 CFR 50.91(a)(1) regarding no significant hazards 
considerations using the standards in 10 CFR 50.92(c). A discussion 
of these standards as they relate to this amendment request follows:
    Criterion 1--Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The ANO-2 Plant Protection System (PPS) includes the electrical 
and mechanical devices and circuitry (from sensors to actuation 
device input terminals) involved in generating signals associated 
with the two protective functions, Engineered Safety Feature 
Actuation System (ESFAS) and Reactor Protective System (RPS). The 
RPS is that portion of the PPS which generates signals that actuate 
a reactor trip. The ESFAS is that portion of the PPS which generates 
signals that actuate Engineered Safety Features (ESF) to mitigate 
the consequences of an accident.
    The ANO-2 Safety Analysis Report (SAR) section 15.1.31 ``Loss Of 
One DC System'' analyzes failure of a DC bus (FODCB) as initiator 
and its causes. The causes for the FODCB are DC leg to leg fault in 
the bus or in the power distribution circuit from the battery. Since 
the proposed change has no impact on the accident initiator, the 
frequency of occurrence is not changed. In order for the FODCB as a 
single failure with an accident to de-energize two [Vital Instrument 
Buses (]VIBs[)], the FODCB would have to occur prior to the safety 
bus

[[Page 66594]]

energization by offsite bus fast transfer or prior to safety bus 
energization by the emergency diesel generator (EDG). The potential 
for de-energization of one pair of VIBs is, therefore, limited to 
time from initiation of the accident to time for safety bus response 
to the secondary plant and Reactor Protective System trips.
    The effects of the FODCB are being revised to assume a secondary 
plant trip that results in de-energization of one power division. 
The existing analysis conclusions remain unchanged. The accident 
analysis is being revised to include de-energization of a pair of 
vital AC instrument channels. De-energization of two vital AC 
sources has not been previously documented as a design bases event.
    Auctioneered bistable power supplies for Plant Protection System 
(PPS) channels A and D are being modified to a single power source 
for each of these two channels. Single channel trips will result for 
all PPS functions in channels A or D for loss of its single channel 
bistable power source. The PPS channels B and C auctioneered power 
supplies remain unchanged to maintain Recirculation Actuation Signal 
(RAS) response to a FODCB.
    Regarding PPS measurement channels with increasing signal 
setpoints, de-energization of a single power supply either results 
in failure of a measurement channel (B or C) to a non-tripped state 
or in failure of a measurement channel (A or D) to a tripped state. 
Neither single channel failure scenario impacts accident initiation 
or mitigation. For PPS measurement channels with decreasing signal 
setpoints the single channel de-energization events result in 
failure of a single affected measurement channel to a tripped state. 
The PPS two out of three logic design with a channel bypassed 
ensures operability with a single channel failure. Neither condition 
impacts accident frequency or consequences.
    With the exception of Recirculation Actuation Signal (RAS) and 
Emergency Feedwater Actuation Signal (EFAS), a FODCB results in an 
automatic ESFAS initiation for those functions with decreasing 
signal setpoints. For other ESFAS functions with a decreasing 
signal, channels A and C or channels B and D fail to the tripped 
state. For those functions with an increasing signal setpoint 
(including EFAS), a FODCB results in a single channel failing not 
tripped, one channel tripping, and two channels remaining 
functional. System level functions remain operable with either a one 
out of two logic (no channels bypassed) or a one out of one logic 
(with a channel bypassed).
    Interposing relay actuation logic has changed from single trip 
path to selective trip path logic. This change insures emergency 
feedwater (EFW) discharge valves will receive an automatic open or 
close demand based on steam generator level and pressure demands.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2--Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    In response to de-energization of a pair of Vital Instrument 
Buses (VIBs), those ESFAS functions with increasing signal 
setpoints, as a minimum, remain functional with one out of one 
logic. One channel trips, one channel does not trip, and two 
channels remain functional. One of the functional channels may be 
bypassed without impact on operability. The trip response of those 
ESFAS functions with decreasing signal to trip setpoints remains 
unchanged.
    EFAS coincidence logic to close the EFW discharge valves 
requires three out of four channels to be in a non-tripped state. 
With a FODCB one channel is tripped, one channel is not tripped, and 
two channels are functional. The close logic becomes two out of two 
with a FODCB.
    By defeating the auctioneered bistable power sources for PPS 
channel A and D bistables, PPS measurement channel A or D will fail 
to its tripped state. This change ensures no more than one channel 
(B or C) fails to a non-tripped state for the FODCB.
    With selective logic EFAS pump discharge valves will receive 
control signals to initiate emergency feedwater and to terminate 
emergency feedwater flow by open and close demands generated 
independent of the 120 Volt channel pair de-energization.
    The existing ANO-2 Failure Modes and Effects Analysis does not 
document failure of a pair of vital instrument AC channels. Neither 
the 120 Volts AC nor the 125 Volt DC system single failure analysis 
assumes failure of two channels of 120 Volts AC. Even though the 
failure of either pair of VIBs caused by a FODCB is not a result of 
the proposed change, the SAR change will address the potential for 
de-energization of a pair of instrument buses. The ANO-2 SAR will be 
updated to reflect the documentation and modification of the PPS 
design to ensure safe plant response.
    Even though the plant response to FODCB is being modified, the 
proposed ANO-2 PPS design resolution does not create the possibility 
of a new or different kind of accident from any previously evaluated 
in the SAR. The PPS will have the redundancy and independence 
sufficient to assure that (1) no single failure results in loss of 
the protection function, and (2) removal from service of any 
component or channel does not result in loss of the required minimum 
redundancy required by the TS. PPS will also meet the single failure 
criterion of IEEE 279-1971 to the extent that any single failure 
within the system does not prevent proper protective action at the 
system level and no single failure will defeat more than one of the 
four protective channels associated with any one trip function.
    Criterion 3--Does Not Involve a Significant Reduction in the 
Margin of Safety.
    Technical Specification Bases 3/4.3.1 & 3/4.3.2 assure 
sufficient PPS redundancy is maintained to permit a channel to be 
bypassed. Under the current design, a FODCB will result in reduction 
of margin by decreasing the number of functional channels to less 
than two. However, with the proposed modification removal from 
service of any component or channel for indefinite bypass will not 
result in loss of the minimum redundancy required by the TS. This 
activity will restore the margin by ensuring ESFAS required 
functions remain capable of automatic actuation with a FODCB.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, Entergy Operations has 
determined that even though the proposed PPS design description 
results in an accident or malfunction of a different type, the 
requested change does not involve a significant hazards 
consideration.

    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: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: John N. Hannon.

Florida Power and Light Company, et al., Docket No. 50-335, St. Lucie 
Plant, Unit No. 1, St. Lucie County, Florida

    Date of amendment request: October 29, 1998.
    Description of amendment request: The proposed amendment would 
revise the terminology used in the St. Lucie Plant Technical 
Specifications (TS) relative to the implementation and automatic 
removal of certain reactor protection system trip bypasses to ensure 
that the meaning of explicit terms used in the TS are consistent with 
the intent of the stated 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:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments are administrative in nature, and do not 
change the function or the setpoints of the RPS trip bypass 
features. The revisions simply make corrections to the Notation of 
TS Tables 2.2-1 and 3.3-1 to ensure that the meaning of explicit 
terms used in the Notes is consistent with the intent of the stated 
requirements based on the St. Lucie plant design. The proposed 
technical specification changes do not involve accident initiators, 
do not change the configuration or method of operation of any plant 
equipment that is used to mitigate

[[Page 66595]]

the consequences of an accident, and do not alter any conditions 
assumed in the plant accident analyses. Therefore, operation of 
either facility in accordance with its proposed amendment would not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendments are administrative in nature and will 
not change the physical plant or the modes of plant operation 
defined in the facility operating licenses. The changes do not 
involve the addition or modification of equipment nor do they alter 
the design or operation of plant systems. Therefore, operation of 
either facility in accordance with its proposed amendment would not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed amendments are administrative in nature and do not 
change the function or the setpoints of the RPS trip bypass 
features. The revisions simply make corrections to the Notation of 
TS Tables 2.2-1 and 3.3-1 to ensure that the meaning of explicit 
terms used in the Notes is consistent with the intent of the stated 
requirements based on the St. Lucie plant design. The proposed 
changes do not alter the basis for any technical specification that 
is related to the establishment of, or the maintenance of, a nuclear 
safety margin. Therefore, operation of either facility in accordance 
with its proposed amendment would not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration. This 
notice is intended to replace an exigent notice of consideration of 
issuance of amendment for St. Lucie Unit 1, previously published as 
exigent TS amendments for both St. Lucie Units 1 and 2 in the Federal 
Register (63 FR 59809). The amendment request for St. Lucie Unit 2 will 
continue to be considered as an exigent amendment as noticed in the 
Federal Register (63 FR 59809).
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Project Director: Frederick J. Hebdon.

GPU Nuclear, Inc, et al., Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

    Date of amendment request: November 10, 1998.
    Description of amendment request: The proposed Technical 
Specification (TS) change would remove the restriction on the sale or 
lease of property within the exclusion area and replace the restriction 
with a requirement to retain complete authority to determine and 
maintain sufficient control of all activities including the authority 
to exclude or remove personnel and property within the minimum 
exclusion distance. A TS Bases page for the proposed change is 
included. Also included are clarifications and administrative changes 
which (1) clarify TS definition 1.38 to become ``Site Boundry'' from 
the current term ``Exclusion Area'' to be consistent with 10 CFR 
20.1003 definition for Site Boundry and the 10 CFR 100.3 definition of 
Exclusion Area, (2) convert the one occurrence of the use of TS 
definition from Exclusion Area to Site Boundry in TS 6.8.4(a)(9), and 
(3) revise and update the Table of Contents for Section I Definitions.`
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would operation of the facility in accordance with the 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    The proposed change is administrative in nature and does not 
affect the purpose, function, performance, operability or testing of 
and does not make any physical or procedural changes to plant 
systems, structures or components. Also, all existing technical 
specification limiting conditions for operation and surveillance 
requirements are retained.
    [Technical Specification Change Request] TSCR 264 does not 
change the size or location of the exclusion area. Since the 
exclusion area size and location are not being changed and no 
physical or procedural changes are being made to the plant, 
radiological consequences in the exclusion area are not affected by 
this TSCR.
    This change addresses the existing technical specification 
restriction on the sale or lease of property within the ``exclusion 
area'' by ensuring that the licensee will retain at all times the 
complete authority to determine and maintain sufficient control of 
all activities through ownership, easement, contract and/or other 
legal instruments on property within the minimum exclusion distance 
including the authority to exclude or remove personnel and property 
within the minimum exclusion distance.
    Therefore, since no physical or procedural changes are being 
made to existing plant systems, structures or components and since 
the proposed change requires the licensee to retain complete 
authority and sufficient control of all activities in the exclusion 
area, operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Would operation of the facility in accordance with the 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    The p[ro]posed change is administrative in nature and does not 
affect the purpose, function, performance, operability or testing of 
and does not make any physical or procedural changes to plant 
systems, structures or components. Also, all existing technical 
specification limiting conditions for operation and surveillance 
requirements are retained.
    This change addresses the existing technical specification 
restriction on the sale or lease of property within the ``exclusion 
area'' by ensuring that the licensee will retain at all times the 
complete authority to determine and maintain sufficient control of 
all activities through ownership, easement, contract and/or other 
legal instruments on property within the minimum exclusion distance 
including the authority to exclude or remove personnel and property 
within the minimum exclusion distance.
    Therefore, since no physical or procedural changes are being 
made to existing plant systems, structures or components and since 
the proposed change requires the licensee to retain complete 
authority and sufficient control of all activities in the exclusion 
area, operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Would operation of the facility in accordance with the 
proposed change involve a significant reduction in a margin of 
safety?
    The p[ro]posed change is administrative in nature and does not 
affect the purpose, function, performance, operability or testing of 
and does not make any physical or procedural changes to plant 
systems, structures or components. Also, all existing technical 
specification limiting conditions for operation and surveillance 
requirements are retained.
    This change addresses the existing technical specification 
restriction on the sale or lease of property within the ``exclusion 
area'' by ensuring that the licensee will retain at all times the 
complete authority to determine and maintain sufficient control of 
all activities through ownership, easement, contract and/or other 
legal instruments on property within the minimum exclusion distance 
including the authority to exclude or remove personnel and property 
within the minimum exclusion distance.
    Therefore, since no physical or procedural changes are being 
made to existing plant

[[Page 66596]]

systems, structures or components and since the proposed change 
requires the licensee to retain complete authority and sufficient 
control of all activities in the exclusion area, operation of the 
facility in accordance with the proposed amendment will not involve 
a significant reduction in margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cecil O. Thomas.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2 (NMP2), Oswego County, New York

    Date of amendment request: October 16, 1998.
    Description of amendment request: The proposed amendment would make 
the following revisions to Technical Specifications (TSs) 3/4.7.1.1: 
(1) Ensure that four service water (SW) pumps are operating with the 
divisional cross connect valves open during Operational Condition 1, 2 
and 3 (current TS requires two SW pumps associated with one loop to be 
operating); (2) Increase the number of division 1 and 2 heaters 
required to be operable from 7 per division per intake to 14 per 
division per intake; (3) The actions necessary for having less than the 
required equipment is being revised to reflect the new limits for SW 
equipment; and (4) SW supply header discharge water temperature is 
being increased from 81 to 82  deg.F. TS 3.7.1.2, Table 3.3.9-1, and 
Table 4.3.9.1-1 are revised to add ``when handling irradiated fuel in 
the secondary containment'' to the applicability section. Table 3.3.9-1 
is being revised to decrease the temperature at which the Intake 
Deicing Heaters are required to be in service from 39 to 38 degrees F. 
TS 3.7.1.2 proposed change is to specify that the necessary portions of 
the SW system needed to support equipment required to be operable shall 
be operable; the Action Section proposed revision reflects this change. 
TS 4.7.1.2.1 surveillance requirement proposed change is to increase 
the flow rate of SW pumps from 6500 GPM to 9000 GPM and to change the 
SW pumps pressure from 80 psi discharge pressure to 70 psi differential 
pressure; TS 4.7.1.2.2 is being revised to decrease the intake tunnel 
water temperature from 39 to 38 degrees F. The surveillance for the 
Intake Deicing Heaters is being changed to reflect the increase in the 
number of heaters required. The title of ``Plant Service Water System'' 
is being changed to ``Service Water System.''
    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 operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The SW System is a once-through system which supplies water from 
Lake Ontario to various essential and non-essential components, as 
required, during normal plant operation and shutdown conditions. The 
System is designed with suitable redundancy to provide a reliable 
source of cooling water for the removal of heat from essential plant 
components, including the RHR [residual heat removal] heat 
exchangers, the EDGs [emergency diesel generators], and room coolers 
for ECCS [emergency core cooling system] equipment, which are 
required for safe reactor shutdown following a LOCA.
    LCO 3.7.1.1 and LCO 3.7.1.2 each currently requires two 
independent SW System loops to be operable, with one of the loops in 
operation. The current LCOs do not provide adequate guidance 
regarding the minimum number of operating pumps. NMPC [Niagara 
Mohawk Power Corporation] proposes to revise LCO 3.7.1.1 and its 
associated Actions and SRs to provide assurance that four SW pumps 
are operable and are operating within acceptable system parameters, 
with the divisional cross-connect valves open, during Operational 
Conditions 1, 2, and 3 to meet the limiting LOCA analysis 
assumptions.
    TS Section 3/4.7.1 currently specifies a maximum SW supply 
header discharge water temperature of 81 degrees F and a limiting 
temperature for Intake Deicing Heater ystem operability (intake 
water) temperature of 39 degrees F. In addition, TS Table 3.3.9-1, 
Action 144, requires the Intake Deicing Heater System heaters to be 
placed in service when the Lake Ontario water temperature reaches 39 
degrees F. NMPC proposes to revise Action 144 of TS Table 3.3.9-1 
and TS LCO 3.7.1.1, including its associated Actions and SRs 
[surveillance requirements], to increase the supply header discharge 
water temperature to its analytical limit of 82 degrees F and reduce 
the limiting temperature for the Intake Deicing Heater System Action 
and operability requirements to 38 degrees F.
    Appropriate changes to LCO 3.7.1.2 and its associated Actions 
and SRs are also proposed in order to assure consistency with the SW 
System analyses assumptions during shutdown conditions. The current 
LCO Actions do not account for the varying flows and heat loads that 
may be required for various plant shutdown conditions. The revision 
to the Applicability for LCO 3.7.1.2 and TS Tables 3.3.9-1 and 
4.3.9.1-1 will assure that the SW System is operable during periods 
when irradiated fuel is being handled in the secondary containment 
and essential loads cooled by the SW System are required to be 
operable (e.g., EDG). A footnote has been added to define 
Operational Condition * and is consistent with similar footnotes in 
the TSs. The proposed changes will assure that the necessary ortions 
of the SW System and the necessary Divisions of the Intake Deicing 
Heater System heaters are operable that are supporting equipment 
required to be operable.
    It is further proposed to change the system title identified in 
the Index and in TS Section 3/4.7.1, including the LCOs and SRs, 
from ``Plant Service Water System'' to ``Service Water System'' to 
be consistent with the NMP2 [Nine Mile Unit 2] UFSAR [Updated Final 
Safety Analysis Report].
    The changes do not involve any physical alteration of the plant, 
and the SW System will remain capable of providing sufficient 
cooling flow for the essential cooling loads during plant operation 
and also during plant shutdown. The changes will have no impact on 
the design or function of the SW System and its components, thus 
assuring that the characteristics and functional performance are 
maintained consistent with the event precursors and the conditions 
and assumptions of the current design basis accident and transient 
analyses. The changes to the LCO AOTs [allowed outage times] are 
either consistent with or are more conservative than the current 
AOTs. Based on the above, adequate assurance is provided that the 
probability of event initiation will remain as previously analyzed. 
Maintaining four pumps operating within acceptable system 
parameters, with the divisional cross connect valves open, during 
Operational Conditions 1, 2, and 3 provides assurance that the 
essential functions supported by the SW System are maintained. 
Particularly, adequate SW flow assures that the primary and 
secondary containments can perform their intended functions of 
limiting the release of radioactive materials to the environment 
following a LOCA. The small (1 degree F) change in the SW supply 
header discharge water (UHS) temperature and Intake Deicing Heater 
System actuation temperature maintain the current design basis for 
the UHS and SW Systems such that there will be no impact on the LOCA 
analyses assumptions or conclusions. The proposed changes to the SW 
System TSs do not adversely affect the capability of plant systems, 
structures, and components to respond to any accident in Operational 
Conditions 4, 5, and *. As a result, there will be no degradation of 
the primary or secondary containment or any other fission product 
barriers which could increase the radiological consequences of an 
accident. In addition, other essential accident mitigation equipment 
supported by the SW System will not be adversely impacted. It is, 
therefore,

[[Page 66597]]

concluded that operation of NMP2, in accordance with the proposed 
amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated. The 
operation of Nine Mile Point Unit 2, in accordance with the proposed 
amendment, will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The changes do not result in any hardware changes or physical 
alteration of the plant which could introduce new equipment failure 
modes, and there will be no impact on the design or function of the 
SW System or its components. The primary and secondary containment 
post-LOCA responses remain within previously assessed limits of 
temperature and pressure. Furthermore, adequate cooling flow is 
assured during plant operation and also during shutdown conditions 
such that essential systems and components remain within their 
applicable design limits. It is, therefore, concluded that no 
requirements are eliminated or new requirements imposed which could 
affect equipment or plant operation such that new credible accidents 
are introduced. Accordingly, operation of NMP2, in accordance with 
the proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The changes provide assurance that the SW System will remain 
capable of providing sufficient cooling flow for the essential 
cooling loads during plant operation and also during plant shutdown 
such that essential systems and components remain within their 
applicable design limits. The changes will have no impact on the 
design or function of the SW System and its components, thus 
assuring that the characteristics and functional performance are 
maintained consistent with the conditions and assumptions of the 
current design basis accident and transient analyses. Maintaining 
four pumps operating within acceptable system parameters, with the 
divisional cross connect valves open, during Operational Conditions 
1, 2, and 3 provides assurance that post-LOCA radioactive releases 
are maintained within 10 CFR 100 limits. The small (1 degree F) 
change in the SW supply header discharge water (UHS) temperature and 
the limiting temperature for the Intake Deicing Heater System Action 
and operability requirements maintains the current design basis for 
the UHS and SW Systems such that there will be no impact on the LOCA 
analyses assumptions or conclusions.
    These changes will not result in a reduction in margin to the 
System analytical limits. Furthermore, maintaining the intake bar 
surface temperature at least 1 degree F above freezing provides an 
adequate margin to prevent the adherence of ice, and provides 
assurance that sufficient flow area is always heated such that the 
SW System will remain capable of providing adequate cooling flow in 
the event of a LOCA. Similarly, maintaining the required SW System 
flow and temperature during Operational Conditions 4, 5, and * will 
assure that the associated equipment is operable such that 
radioactive releases are maintained within 10 CFR 100 limits. It is, 
therefore, concluded that the changes do not eliminate any 
requirements, impose any new requirements, or alter any physical 
parameters which significantly reduce the margin to an acceptance 
limit or adversely affect the margins associated with the fission 
product barriers as established by the design basis accident and 
transient analyses. Accordingly, operation of NMP2, in accordance 
with the proposed amendment, will not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: S. Singh Bajwa.

Northeast Nuclear Energy Company (NNECO) et al., Docket No. 50-336, 
Millstone Nuclear Power Station, Unit No. 2, New London County, 
Connecticut

    Date of amendment request: September 28, 1998.
    Description of amendment request: The proposed amendment would 
change Technical Specifications 3.3.2.1, ``Instrumentation--Engineered 
Safety Features Actuation System''; 3.4.6.2, ``Reactor Coolant System--
Reactor Coolant System Leakage''; 3.4.8, ``Reactor Coolant System--
Specific Activity''; 3.6.2.1, ``Containment Systems--Depressurization 
and Cooling Systems Containment Spray and Cooling Systems''; 3.6.5.1, 
``Containment Systems--Secondary Containment Enclosure Building 
Filtration System''; 3.7.6.1, ``Plant Systems--Control Room Emergency 
Ventilation System''; and 3.9.15, ``Refueling Operations--Storage Pool 
Area Ventilation System--Fuel Storage.'' Information would also be 
added to the Bases of the associated Technical Specifications to 
address the proposed changes.
    The proposed amendment would also revise the Operating License DPR-
65 by incorporating a change to the Millstone Unit No. 2 Final Safety 
Analysis Report (FSAR). The change to the FSAR is associated with the 
revised main steamline break analyses, new determination of the 
radiological consequences of a main steamline break, and a revised 
determination of the radiological consequences of the design basis 
loss-of-coolant accidents (LOCAs).
    The proposed changes to the main steamline break analysis, as 
described in the FSAR, are based on the revised Siemens Power 
Corporation steamline break methodology. The report describing the 
revised methodology was submitted by Siemens Power Corporation to the 
NRC for approval in a letter dated June 30, 1998. The revised 
methodology was used to perform the Millstone Unit No. 2 plant-specific 
analysis for post-scram main steamline break. This plant-specific 
analysis was submitted by NNECO in a letter dated August 12, 1998, 
which proposed to change the list of documents in the Technical 
Specifications that describe the analytical methods used to determine 
the core operating limits. The proposed changes contained in this 
letter assume approval of the previously submitted revised Siemens 
Power Corporation steamline break methodology, and the changes to the 
list of documents in the Millstone Unit No. 2 Technical Specifications 
that describe the analytical methods used to determine the core 
operating limits.
    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 10CFR50.92, NNECO 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 10CFR50.92(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.

Analyses Changes

    The main steam line break analyses and the determinations of the 
radiological consequences of the main steam line break and loss of 
coolant accident have been revised. A brief summary of the 
significant changes to the main steam line break analyses and the 
radiological consequences of the main steam line break and loss of 
coolant accident is presented below.
    1. The limited fuel failure following a main steam line break 
outside containment results in an increase in the calculated 
radiological consequences both off-site and in the control room. To 
limit the consequences of a main steam line break outside 
containment, the

[[Page 66598]]

Technical Specification allowed steam generator tube leakage will be 
reduced to 0.035 gpm [gallons per minute] per steam generator.
    2. Credit will now be taken for iodine removal from the 
containment atmosphere by the Containment Spray System (CSS). The 
use of the CSS for iodine removal has not been previously approved 
by the NRC.
    3. The proposed increase to the allowable control room in-
leakage will provide additional operational flexibility to address 
expected minor system degradation over time. The increase in the 
allowable control room in-leakage will result in an increase in the 
calculated dose to the Control Room Operators.
    4. The addition of the dose consequences from containment sump 
backleakage to the Refueling Water Storage Tank (RWST) has been 
included in the off-site and control room loss of coolant accident 
(LOCA) analyses increases the consequences of previously evaluated 
accidents.
    The containment sump backleakage into the RWST results in sump 
water entering the RWST when the RWST is at its minimum level. The 
RWST will become a radioactive source and contribute a shine dose to 
the surrounding areas. The increase in dose rates onsite will not 
prevent operators from remaining in the control room or from 
accessing equipment needed to mitigate the accident.
    All piping and valves associated with RWST backleakage are 
located in a harsh radiation area. Backflow from the sump might 
increase dose rates in the area where these components are located. 
Additional dose contributions, where they occur, do not adversely 
impact the environmental qualification of the vital equipment 
located there. All vital equipment would continue to perform its 
safety function.
    5. Credit will be taken in the main steam line break analyses 
for the recently installed cavitating venturis in the Auxiliary 
Feedwater System. However, this will not change the amount of fuel 
failure. Therefore, credit for this equipment will not impact the 
radiological consequences of a main steam line break.
    6. Credit will be taken for the Reactor Coolant System (RCS) low 
flow reactor trip for the pre-scram inside containment main steam 
line break analysis. This equipment will be qualified for the 
expected containment environment following a main steam line break 
inside containment and will be added to the Environmental 
Qualification Master List.
    7. Millstone Unit No. 1 design basis accidents, loss of coolant 
and main steam line break, will no longer be evaluated for impact on 
Millstone Unit No. 2 control room habitability. This credits the 
decision to decommission Millstone Unit No. 1. [Footnote--B.D. 
Kenyon letter to the NRC, ``Millstone Nuclear Power Station, Unit 
No. 1 Certification of Permanent Cessation of Power Operations and 
that Fuel Has Been Permanently Removed from the Reactor,'' dated 
July 21, 1998.]
    The revised main steam line break analyses and the revised 
determinations of the radiological consequences of the main steam 
line break and design basis LOCA analyses take credit for equipment 
not previously assumed in the analyses, and for plant or equipment 
operating restrictions not currently contained in the Technical 
Specifications. The changes to the analyses will not adversely 
affect the probability of an accident previously evaluated, but the 
revised analyses results do indicate that the consequences of an 
accident previously evaluated will increase. Specifically, the 
following changes cause an increase in the consequences of an 
accident previously evaluated.
    1. The increase in allowable control room in-leakage from 100 
SCFM [standard cubic feet per minute] to 130 SCFM when the Control 
Room Emergency Ventilation System is operating in the recirculation/
filtration mode.
    The dose to the Control Room Operators from a Millstone Unit No. 
2 LOCA increased from 9.25 to 25.8 rem to the thyroid and from 0.205 
to 2.29 rem to the skin. The dose to the whole body decreased. (Both 
low wind speed and high wind speed release conditions were analyzed. 
The low wind speed condition bounds the high wind speed condition.) 
The dose to the Control Room Operators from a Millstone Unit No. 3 
LOCA increased from 2.67 to 14 rem to the skin and from 0.209 to 
1.484 rem to the whole body. The dose to the thyroid decreased. The 
doses to the Control Room Operators from either a Millstone Unit No. 
2 or Unit No. 3 LOCA remain below the GDC [General Design Criterion] 
19 criteria of 30 rem thyroid, 5 rem whole body and 30 rem to the 
skin.
    The new calculated doses to the Millstone Unit No. 2 Control 
Room Operators from a main steam line break outside containment are 
29 rem thyroid, 0.03 rem whole body and 0.5 rem skin. The doses to 
the Millstone Unit No. 2 Control Room Operators are below the GDC 19 
criteria of 30 rem thyroid, 5 rem whole body, and 30 rem to the 
skin. (Note: The dose to the Control Room Operators from a main 
steam line break was not previously evaluated because fuel failure 
was not predicted to occur.)
    2. The limited fuel failure that is predicted in the revised 
main steam line break analyses.
    Previously, the radiological consequences of a main steam line 
break were not determined and were not presented in the FSAR because 
fuel failure was not predicted to occur. Because of the predicted 
limited fuel failure for the main steam line break outside of 
containment, the radiological consequences were analyzed. The 
results to the Exclusion Area Boundary (EAB) are 4.8 rem thyroid and 
0.06 rem whole body. The results to the Low Population Zone (LPZ) 
are 2.3 rem thyroid and 0.02 rem whole body. To meet the dose 
acceptance criteria to the Millstone Unit No. 2 Control Room 
Operators, the maximum allowable Technical Specification primary to 
secondary leak rate is being reduced to 0.035 gpm per steam 
generator. The results to the Millstone Unit No. 2 Control Room 
Operators are 29 rem thyroid, 0.03 rem whole body and 0.5 rem skin. 
The main steam line break outside containment is the limiting 
accident for the Millstone Unit No. 2 Control Room Operators. 
However, the dose consequences of a main steam line break are less 
than the 10CFR100 limits off-site of 300 rem thyroid and 25 rem 
whole body, and the doses to the Millstone Unit No. 2 Control Room 
Operators are below the GDC 19 criteria of 30 rem thyroid, 5 rem 
whole body, and 30 rem to the skin.
    3. Taking credit for the low RCS flow reactor trip for the pre-
scram inside containment main steam line break analysis.
    Previous analyses did not credit the low RCS flow reactor trip 
in a harsh environment. This credits the low flow trip in a manner 
not previously reviewed by the NRC for Millstone Unit No. 2. Without 
credit for this reactor trip, the predicted fuel failure for steam 
line breaks inside containment would be higher.
    4. Taking credit for the removal of radioactive iodine from the 
containment atmosphere by containment spray.
    Previous analyses did not rely on the spray function to reduce 
iodine concentration in the post-accident atmosphere inside 
containment. This adds a mitigation function to the CSS that has not 
been previously reviewed by the NRC for Millstone Unit No. 2. 
Without credit for the removal of iodine, the predicted dose 
consequences following a LOCA would be higher.
    5. The addition of sump backleakage to the RWST during a LOCA.
    The resultant dose contribution to the LPZ from RWST backleakage 
is 1.487 rem thyroid and 0.11 rem whole body. The total dose to the 
LPZ from a design basis LOCA is 21.86 rem thyroid and 0.941 rem 
whole body. The dose is well below the 10CFR100 limits of 300 rem 
thyroid and 25 rem whole body. The dose to the EAB was not affected 
because leakage into the RWST does not start until 25.45 hours post-
LOCA and the EAB is a 2-hour dose.
    The resultant dose contribution to the Millstone Unit No. 2 
Control Room Operators from RWST backleakage is 3.75 rem thyroid, 
0.017 rem whole body and 0.296 to the skin. The total dose to the 
Millstone Unit No. 2 Control Room Operators from the LOCA is 25.8 
rem thyroid, 0.718 rem whole body and 2.29 rem to the skin. These 
doses are below the GDC 19 limits of 30 rem thyroid and skin, and 5 
rem whole body.
    The analyses results meet the guidance contained in SRP 
[Standard Review Plan] 15.1.5, SRP 15.6.5, and the limits of 
10CFR100 and GDC 19. Therefore, there will be no significant 
increase in the probability or consequences of an accident 
previously evaluated.

Technical Specification Changes

    Technical Specification Non-Technical Changes
    The minor editorial and non-technical changes to correct 
spelling (Technical Specification 3.3.2.1), modify the title of a 
table column (Technical Specification 3.4.8), clarify the type of 
measurement performed (Technical Specification 3.4.8), and establish 
consistent terminology (Technical Specification 3.7.6.1) will not 
result in any technical changes to the Millstone Unit No. 2 
Technical Specifications. The proposed changes will have no adverse 
effect on plant

[[Page 66599]]

operation. Therefore, there will be no significant increase in the 
probability or consequences of an accident previously evaluated.

Technical Specification 3.4.6.2

    The reduction in the maximum allowable value of primary to 
secondary leakage per steam generator is consistent with the new 
radiological assessment of the potential control room operator 
exposure following a main steam line break outside of containment. 
The wording change to SR [Surveillance Requirement] 4.4.6.2.1 will 
clarify that the water inventory balance is used to verify 
compliance with the identified and unidentified leakage limits. 
Pressure boundary leakage would first show up as unidentified 
leakage during performance of SR 4.4.6.2.1. Further investigation, 
(plant walkdown) would be necessary to classify the unidentified 
leakage as pressure boundary leakage. This is consistent with 
established plant practices to detect pressure boundary leakage.
    The addition of the new SR 4.4.6.2.2 will address the primary to 
secondary leakage limit. The new SR will include an exception to 
Technical Specification 4.0.4 that will allow the determination of 
primary to secondary leakage to be deferred until after Mode 4 is 
entered. Even though verification of compliance with the primary to 
secondary limit will not be done prior to entering Mode 4, the limit 
is still expected to be met.
    The proposed changes will have no adverse effect on plant 
operation. Therefore, there will be no significant increase in the 
probability or consequences of an accident previously evaluated.

Technical Specification 3.4.8

    The addition of the words ``of gross specific activity'' to the 
Limiting Condition for Operation (LCO), Action Statements, and SR 
will clarify what the E-Bar limit applies to. This is consistent 
with the Technical Specification Definition (1.20) for E-Bar.
    The addition of a footnote (*) to state the power history 
requirements for the determination of E-Bar will ensure that the 
necessary plant conditions are established prior to performing the 
analysis. This will not affect the E-Bar LCO limit or the 
requirement to perform the analysis. The proposed change is 
consistent with NUREG--0212 and NUREG--1432.
    The footnote will also specify that the provisions of 
Specification 4.0.4 are not applicable. This will allow entry into 
Mode 1, without determining the value of E-Bar, assuming that the 
power history requirements will not be met until after Mode 1 is 
entered. This will normally only apply following an extended 
shutdown.
    The Isotopic Analysis for Iodine (including I-131, I-133, and I-
135) sample requirement will be expanded to include the LCO 
requirement for 100/E-Bar. This is consistent with the requirements 
of Action Statement d. This change will expand the sampling 
requirement for iodine. Minor wording changes will also be made to 
be consistent with the proposed changes to the LCO wording.
    The proposed changes will have no adverse effect on plant 
operation. Therefore, there will be no significant increase in the 
probability or consequences of an accident previously evaluated.

Technical Specification 3.6.2.1

    The revised radiological assessment calculation for the design 
basis accident credits iodine removal from the containment 
atmosphere by the CSS. This will require a reduction in the allowed 
outage time (AOT) of one containment spray train from seven days to 
seventy two hours. This AOT is consistent with NUREG-0212 and NUREG-
1432. This will help ensure that plant equipment assumed in the 
safety analyses will be available. This is a more restrictive change 
which will have no adverse effect on plant operation. Therefore, 
there will be no significant increase in the probability or 
consequences of an accident previously evaluated.

Technical Specification 3.6.5.1

    The value for the pressure drop across the combined HEPA [high-
efficiency particulate air] filters and charcoal adsorber banks 
specified in SR 4.6.5.1.d.1 will be changed from a generic value 
[less than or equal to] 6 inches water gauge) to a plant specific 
value [less than or equal to] 2.6 inches water gauge). This is a 
more restrictive change which will have no adverse effect on plant 
operation. Therefore, there will be no significant increase in the 
probability or consequences of an accident previously evaluated.

Technical Specification 3.7.6.1

    The value for the pressure drop across the combined HEPA filters 
and charcoal adsorber banks specified in SR 4.7.6.1.e.1 will be 
changed from a generic value [less than or equal to] 6 inches water 
gauge) to a plant specific value [less than or equal to] 3.4 inches 
water gauge). This is a more restrictive change which will have no 
adverse effect on plant operation.
    SR 4.7.6.1.e.2 will be expanded to clarify that the test of the 
capability of the Control Room Emergency Ventilation Trains to 
switch to the recirculation mode is performed with the trains 
initially operating in the normal mode and the smoke purge mode of 
operation. This will not affect the requirement that the trains be 
capable of switching to the recirculation mode.
    The value of allowable control room air in-leakage specified in 
SR 4.7.6.1.e.3 will be increased from 100 SCFM to 130 SCFM. This is 
consistent with the recently revised control room radiological 
analysis for the design basis accidents.
    The proposed increase will provide additional operational 
flexibility to address expected minor system degradation over time. 
This increase is supported by the new analysis.
    The proposed changes will have no adverse effect on plant 
operation. Therefore, there will be no significant increase in the 
probability or consequences of an accident previously evaluated.

Technical Specification 3.9.15

    The value for the pressure drop across the combined HEPA filters 
and charcoal adsorber banks specified in SR 4.9.15.d.1 will be 
changed from a generic value [less than or equal to] 6 inches water 
gauge) to a plant specific value [less than or equal to] 2.6 inches 
water gauge). This is a more restrictive change which will have no 
adverse effect on plant operation. Therefore, there will be no 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes 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 changes are consistent with the revised 
analyses, and the appropriate acceptance criteria are met for the 
revised analyses. Therefore, the license amendment request does not 
impact the probability of an accident previously evaluated nor does 
it involve a significant increase in the consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes will not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. They do not alter the way any 
structure, system, or component functions and do not alter the 
manner in which the plant is operated. The proposed changes do not 
introduce any new failure modes.
    Also, the response of the plant and the operators following 
these accidents is unaffected by the change. Therefore, the proposed 
changes 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.

Analyses Changes

    The acceptance criteria for a main steam line break in the SRP 
15.1.5 does not exclude the prediction of fuel failure. Instead, the 
SRP requires that ``Any fuel damage calculated to occur must be of 
sufficiently limited extent that the core will remain in place and 
intact with no loss of core cooling.'' The limited fuel failure that 
is now predicted in the revised main steam line break analyses meets 
this acceptance criterion. In addition, the RCS low flow reactor 
trip that is now being credited to function in a harsh environment 
to limit fuel failure is already required to be operable by 
Technical Specifications.
    The revised dose consequences for the design basis accidents 
assumes a control room in-leakage of 130 SCFM. In addition, iodine 
removal by the CSS, which is already required to be operable by 
Technical Specifications, is assumed. The acceptance criteria for 
the dose consequences of the design basis accidents to the EAB, LPZ 
and the control room personnel is met in the revised analyses. 
Therefore, the revisions to the dose consequence analyses for the 
design basis accidents do not involve a significant reduction in the 
margin of safety.

[[Page 66600]]

Technical Specification Changes

    The proposed changes will correct spelling and terminology 
errors, reduce the maximum allowable primary to secondary leakage, 
add a new surveillance requirement, modify surveillance requirements 
for RCS specific activity, reduce the allowed outage time for a 
containment spray train, reduce the allowed pressure drop across the 
control room and enclosure building HEPA [high-efficiency 
particulate air] filters, and increase the control room maximum 
allowed in-leakage. These changes will have no adverse effect on 
equipment important to safety. The equipment will continue to 
function as assumed in the design basis accident analysis. 
Therefore, there will be no significant reduction of the margin of 
safety as defined in the Bases for the Technical Specifications 
affected by these proposed changes.
    The only adverse impact of the proposed changes is that the dose 
consequences following an accident may increase. However, the 
revised analyses show that the acceptance criteria for the accident 
analyses are met. Therefore, based on the responses above, the 
proposed changes are deemed safe.
    The NRC has provided guidance concerning the application of 
standards in 10CFR50.92 by providing certain examples (March 6, 
1986, 51 FR 7751) of amendments that are considered not likely to 
involve an SHC. The minor editorial and non-technical changes 
proposed herein to correct reference, spelling, and terminology 
errors are enveloped by example (i), a purely administrative change 
to Technical Specifications. The changes proposed herein to add a 
new surveillance requirement to verify primary to secondary leakage 
and to reduce the allowable pressure drop across various ventilation 
filters are enveloped by example (ii), a change that constitutes an 
additional limitation, restriction, or control not presently 
included in the Technical Specifications. All of the other changes 
proposed herein are not enveloped by any specific example.
    As described above, this License Amendment Request does not 
impact the probability of an accident previously evaluated, does not 
involve a significant increase in the consequences of an accident 
previously evaluated, does not create the possibility of a new or 
different kind of accident from any accident previously evaluated, 
and does not result in a significant reduction in a margin of 
safety. Therefore, NNECO has concluded that the proposed changes do 
not involve an SHC.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Project Director: William M. Dean.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: October 22, 1998.
    Description of amendment request: The licensee is proposing to 
change Technical Specifications 3.3.2.1, ``Instrumentation--Engineered 
Safety Feature Actuation System Instrumentation''; 3.4.9.3, ``Reactor 
Coolant System [RCS]--Overpressure Protection Systems''; and 3.5.3, 
``Emergency Core Cooling Systems--ECCS Subsystems--Tavg < 300 [degrees] 
F.'' The proposed changes will allow Millstone Unit No. 2 to prevent an 
automatic start of any high-pressure safety injection (HPSI) pump when 
the shutdown cooling system (SDCS) is in operation (Mode 4 and below). 
An inadvertent start of an HPSI pump could result in overpressurization 
of the SDCS.
    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 10CFR50.92, Northeast Nuclear Energy Company 
(NNECO) 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 
10CFR50.92(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 changes to Technical Specifications 3.3.2.1 and 
3.5.3 will no longer require the HPSI pump, required to be operable 
in Mode 4, to start automatically on a Safety Injection Actuation 
Signal (SIAS). (The automatic SIASs on low pressurizer pressure and 
high containment pressure are not required to be operable in Mode 4. 
However, the manual safety injection pushbuttons are required in 
Mode 4). This will allow the operable HPSI pump control switch to be 
placed in the pull-to-lock position without affecting the 
operability of that pump. All HPSI pumps will be prevented from 
automatically starting when the plant is in Mode 4, and the Shutdown 
Cooling System (SDCS) is aligned to the RCS to prevent an 
inadvertent start of a[n] HPSI pump which could overpressurize the 
SDCS. These changes will not reduce the requirement for at least one 
HPSI pump to be operable in Mode 4. The changes will require an 
additional operator action to remove the operable HPSI pump breaker 
control switch from the pull-to-lock position, in addition to 
initiating safety injection by use of the manual pushbuttons, if 
Safety Injection System actuation is needed in Mode 4. The 
requirement to manually initiate a[n] HPSI pump, in addition to 
manually initiating a[n] SIAS, does not involve complicated 
equipment manipulations nor require extensive time for performing 
the required operator actions. The HPSI pump control switches are 
located in the Control Room on the same panels as the manual SIAS 
pushbuttons. The additional step required to start a[n] HPSI pump 
will not add any appreciable time for initiating HPSI flow while in 
Mode 4. In addition, considering the lower probability of a 
significant loss of coolant accident in Mode 4, and the slower plant 
response to a loss of coolant accident in Mode 4, the time required 
for the additional operator action will have no significant effect 
on the consequences of the accident. Therefore, there will be no 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change to Technical Specification 3.4.9.3, 
Surveillance Requirement (SR) 4.4.9.3.3, will allow the use of the 
new pull-to-lock feature of the HPSI pump control switches to 
satisfy low temperature overpressure protection mass input 
requirements. This will not affect either the LTOP [low-temperature 
overpressure protection] HPSI pump mass input restrictions or the 
level of control to ensure the HPSI pumps are not capable of 
injecting into the RCS. The proposed changes will have no adverse 
effect on plant operation. Therefore, there will be no significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed minor editorial and non-technical changes to add 
amendment numbers to Page 3/4 3-12 and to revise the wording of SRs 
4.4.9.3.2 and 4.4.9.3.3 will not result in any technical changes to 
the Millstone Unit No. 2 Technical Specifications. The proposed 
changes will have no adverse effect on plant operation. Therefore, 
there will be no significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to the Bases reflect the proposed changes 
to the applicable Technical Specifications. The proposed changes 
will have no adverse effect on plant operation. Therefore, there 
will be no significant increase in the probability or consequences 
of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes will allow the use of the HPSI pump breaker 
control switch

[[Page 66601]]

pull-to-lock feature. Operation of the HPSI pump in Mode 4 will 
change since the operator will have to start the HPSI pump, in 
addition to manually initiating safety injection. However, HPSI pump 
operation is not an accident initiator. Therefore, the proposed 
changes 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 Technical Specification changes will no longer 
require the HPSI pump, required to be operable in Mode 4, to start 
automatically on a[n] SIAS, will allow the use of the new pull-to-
lock feature of the HPSI pump control switches to satisfy low 
temperature overpressure protection mass input requirements, and 
will make minor editorial and non-technical changes. These changes 
will have no adverse effect on equipment important to safety. The 
equipment will continue to function as assumed in the design basis 
accident analysis. Therefore, there will be no significant reduction 
in the margin of safety as defined in the Bases for the Technical 
Specifications affected by these proposed changes.
    The only adverse impact of the proposed changes is that an 
additional operator action will be necessary to initiate HPSI flow 
in Mode 4, if needed. However, considering the lower probability of 
a significant loss of coolant accident in Mode 4, and the slower 
plant response to a loss of coolant accident in Mode 4, the time 
required for the additional operator action will have no significant 
effect on the consequences of the accident. Therefore, based on the 
responses above, the proposed changes are deemed safe.
    The NRC has provided guidance concerning the application of 
standards in 10CFR50.92 by providing certain examples (March 6, 
1986, 51 FR 7751) of amendments that are considered not likely to 
involve an SHC. The minor editorial and non-technical changes 
proposed herein to add page amendment numbers and clarify wording 
are enveloped by example (i), a purely administrative change to 
Technical Specifications. All of the other changes proposed herein 
are not enveloped by any specific example.
    As described above, this License Amendment Request does not 
impact the probability of an accident previously evaluated, does not 
involve a significant increase in the consequences of an accident 
previously evaluated, does not create the possibility of a new or 
different kind of accident from any accident previously evaluated, 
and does not result in a significant reduction in a margin of 
safety. Therefore, NNECO has concluded that the proposed changes do 
not involve an SHC.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Project Director: William M. Dean.

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating 
Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: October 30, 1998.
    Description of amendment request: Limerick Generating Station 
(LGS), Units 1 and 2, Technical Specifications (TS) Surveillance 
Requirements 4.8.4.3.b.1, 4.8.4.3.b.2, and 4.8.4.3.b.3 list the 
Overvoltage (OV), Undervoltage (UV), and Underfrequency (UF) values for 
the protective instrumentation for the RPS electric power monitoring 
channels. The proposed changes correct a discrepancy between the 
General Electric Nuclear Engineering (GENE) Design Specification for 
Power Supply Monitoring Relays and the existing TS Allowable Values 
(AVs). The changes will revise the OV, US, and UF values from 132VAC, 
109VAC, and 57Hz to 127.6VAC, 110.7VAC, and 57.05Hz respectively.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed Tech Spec changes to section 4.8.4.3.b for the 
Overvoltage (OV), Undervoltage (UV), and Underfrequency (UF) relays 
are more conservative than the existing TS values. This change 
provides more protection for the associated RPS components, thus 
decreasing the probability of a failure in RPS. The associated Non-
Conformance Report and calculation provide assurance that the OV/UV/
UF settings are acceptable since the calculated values assure that 
the RPS components will operate within their ratings. There are no 
physical changes to the associated protective relays by the TS 
change; thus, original design basis redundancy and separation is 
maintained. There is no change in the interface of the RPS and its 
power supplies.
    The safety function of the RPS is to initiate a reactor scram in 
order to protect the primary fission products barrier, the reactor 
fuel. The proposed TS Change to impose more conservative Allowable 
Values for the OV, UV, and UF relays will provide additional 
assurance that the RPS will operate within equipment voltage and 
frequency ratings, and will not be damaged by power system 
anomalies. This change will not affect the scram function of RPS; 
thus, the consequences of any design basis events will not be 
affected.
    Therefore, the proposed TS changes do not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS Allowable Values changes will not result in any 
physical changes to the RPS Electric Power Monitoring System. 
Existing setpoints will not be changed, only the TS Allowable Values 
are being modified to be more conservative.
    The system redundancy and independence are not changed, and no 
new failure modes are introduced.
    Therefore, the proposed TS changes do not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    Currently, there are no TS bases for the existing RPS Electric 
Power Monitoring System OV, UV, and UF allowable values. Specific 
analytical limits for system voltage and frequency are not defined 
in the Safety Analysis Report, nor discussed in any design basis 
Allowed Outage Time or accident evaluation.
    Investigation into the licensing basis has identified nominal 
values of +/-10% of 120 VAC and -5% of 60 HZ for the Allowable 
Values. These values are included in NUREG 0123, from which LGS's 
TSs were developed. NUREG 0123 also provides no bases for these 
values.
    The proposed changes in the TS Allowable Values is based on a 
revision to the calculation for RPS Breaker Panel--RPS / UPS 
[uninterruptible power supply] System Bus Relay Settings. This 
revision determines the new allowable values based on the design 
ratings of RPS components, and factors in instrument inaccuracies 
and margin. These changes will also provide bases for the associated 
TS section. The proposed changes bring TSs into agreement with plant 
design specifications.
    Therefore, the proposed TS changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

[[Page 66602]]

    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.
    Attorney for licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101.
    NRC Project Director: Robert A. Capra.

Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek 
Generating Station, Salem County, New Jersey

    Date of amendment request: October 22, 1998.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 4.8.2.1.b.3 to increase the minimum 
battery electrolyte temperature limit from 60 deg.F to 72 deg.F. This 
change resolves a discrepancy in the electrolyte temperature assumed in 
the Class 1-E battery sizing calculations versus the limit specified in 
the TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed TS change does not involve any physical changes to 
plant structures, systems or components (SSC). The Class-1E 
batteries will continue to function as designed. The Class-1E 
battery system is designed to mitigate the consequences of an 
accident, and therefore, can not contribute to the initiation of any 
accident. The proposed TS surveillance testing and monitoring 
requirements will continue to ensure that the Class-1E batteries are 
capable of performing their required safety functions. In addition, 
this proposed TS change will not increase the probability of 
occurrence of a malfunction of any plant equipment important to 
safety, since the manner i[n] which the Class-1E battery system is 
operated is not affected by these proposed changes. The proposed 
changes merely establish TS surveillance acceptance criteria that 
more appropriately reflect the actual plant design. Therefore, the 
proposed TS changes would not result in an increase of the 
consequences of an accident previously evaluated.
    Therefore, the proposed TS change does not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    (2) The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed TS changes do not involve any physical changes to 
the design of plant systems, structures or components. The design 
and operation of the Class-1E battery system is not changed from 
that currently described in the [Updated Final Safety Analysis 
Report] UFSAR, only the allocation of battery capacity design margin 
is affected by the increased TS minimum battery electrolyte 
temperature limit. The Class-1E battery system will continue to 
function as designed to mitigate the consequences of an accident. 
Implementing new TS surveillance acceptance criteria that more 
appropriately reflect the actual plant design does not permit plant 
operation in a configuration that would create a different type of 
malfunction to the Class-1E batteries than any previously evaluated. 
In addition, the proposed TS changes do not alter the conclusions 
described in the UFSAR regarding the safety related functions of the 
Class-1E batteries or their support systems.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    (3) The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed TS change involves the implementation of new TS 
surveillance acceptance criteria that more appropriately reflect the 
actual plant design. The new TS minimum battery electrolyte 
temperature limit enables the Class-1E battery capacity margin to be 
allocated in a manner which conforms to Hope Creek's current 
licensing basis. The ability of the Class-1E batteries to 
independently supply their required loads for four hours without 
support from battery chargers is not affected by these proposed 
changes. The safety-related Class-1E support systems will ensure 
that the proposed TS minimum electrolyte temperature limit is met.
    Therefore, the proposed TS change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: Robert A. Capra.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of amendment request: October 15, 1998, as supplemented by 
letter dated November 11, 1998.
    Description of amendment request: The proposed amendments would 
change the Vogtle Electric Generating Plant, Unit 1 and Unit 2 Facility 
Operating Licenses to delete or modify certain license conditions, 
which have become obsolete or inappropriate. In addition, the Technical 
Specifications would be reconstituted to reflect revised word 
processing. No change in technical requirements would be involved; 
however, the font would be changed to Arial 11 point; page numbers 
would be revised to a limiting condition for operation specific 
numbering scheme; and intentional blank pages would be deleted.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes either remove or modify provisions in the 
VEGP [Vogtle Electric Generating Plant] Unit 1 and [Unit] 2 
Operating Licenses that have been completed or are otherwise 
obsolete. Each proposed change is summarized below:
    Certain Surveillance Requirements (SRs) that were either added 
or modified at the time of Improved Technical Specifications (ITS) 
implementation were listed in the Operating Licenses with a schedule 
for performance. With the exception of Unit 2 SR 3.8.1.20, all SRs 
are deleted from the Operating Licenses, because they have since 
been performed according to schedule, and will henceforth be 
performed in accordance with the Technical Specifications.
    A condition concerning changes to the Unit 1 initial test 
program is deleted due to the completion of the program.
    A condition related to FEMA [Federal Emergency Management 
Agency] procedures and the emergency plan is deleted from the Unit 1 
license due to the obsolescence of the condition.
    Conditions requiring the submission of Unit 1 reports concerning 
the steam generator tube rupture analysis, the reactor vessel level 
instrumentation system, the safety parameter display system, the 
detailed control room design review, and the zinc coating of the 
diesel fuel storage tanks are deleted due to completion of the 
required activities.
    A condition requiring modification of the Unit 1 ventilation 
exhaust of the alternate radwaste facility is deleted due to 
completion of the required activity.
    An exemption related to the seismic adequacy of the Unit 1 spent 
fuel racks is deleted because the required actions are completed and 
the exemption has been determined to be no longer in effect.
    A condition in both the Unit 1 and Unit 2 licenses containing 
reporting requirements for other license conditions is revised due 
to ambiguities between the requirements in the license condition and 
those published in NRC regulations.
    A schedular exemption for the Unit 2 decommissioning funding 
report is deleted

[[Page 66603]]

because the report was submitted as required and the exemption is no 
longer in effect.
    The Technical Specifications and associated Bases have been 
converted from WordPerfect for DOS version 5.1 to 
Microsoft Word 97. There were no changes to technical 
requirements. The only visible changes to the document are as 
follows: (1) the font was changed to Arial 11 point; [(2)] page 
numbers were revised to an LCO [limiting condition for operation] 
specific numbering scheme; and [(3)] intentionally blank pages were 
deleted.
    The proposed changes discussed above are strictly 
administrative/editorial and do not affect the operation or function 
of any plant system, component, or structure. Therefore, the 
proposed changes do not increase the probability of occurrence or 
the consequences of a previously evaluated accident.
    2. The proposed changes do not create the possibility of a new 
and different type of accident from any previously evaluated.
    The proposed administrative/editorial changes do not alter the 
operation of any plant system or equipment and do not introduce a 
new mode of operation. Each requirement contained in the license 
conditions proposed for deletion has either been completed or is 
obsolete. Since these parts of the license are no longer applicable, 
deletion of these items does not provide the potential for an 
accident to be created. The conversion of the Technical 
Specifications from one word processing format to another did not 
involve any changes to technical requirements. Thus, the proposed 
changes cannot create a new accident initiating mechanism, and do 
not create the possibility of a new and different type of accident 
from any previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The license conditions proposed for deletion are obsolete and 
each requirement has been completed. The conversion of the Technical 
Specifications from one word processing format to another did not 
involve any changes to technical requirements. Since the proposed 
changes are strictly administrative/editorial and do not involve any 
physical or procedural changes to the plant, the margin of safety, 
as defined in the bases for any Technical Specification is not 
affected by the proposed changes.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia.
    NRC Project Director: Herbert N. Berkow.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: November 16, 1996 (TS 98-06).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah Nuclear Plant Technical Specifications (TSs) by 
revising the emergency diesel generator (EDG) surveillance requirements 
(SRs) to add a note that allows the SR to be performed in Modes 1, 2, 3 
or 4, if the associated components are already out-of-service for 
testing or maintenance and to remove the SR that verifies certain 
lockout features prevent EDG starting.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:

    TVA has concluded that operation of SQN Units 1 and 2, in 
accordance with the proposed change to the TSs, does not involve a 
significant hazards consideration. TVA's conclusion is based on its 
evaluation, in accordance with 10 CFR 50.91(a)(1), of the three 
standards set forth in 10 CFR 50.92(c).
    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The probability of occurrence or the consequences for an 
accident or malfunction of equipment is not increased by this 
request. The proposal does not alter the way any structure, system 
or component functions, does not modify the manner in which the 
plant is operated, and does not alter equipment out-of-service time. 
This request does not degrade the ability of the D/G [emergency 
diesel generator] or equipment downstream of the load sequencers to 
perform their intended function. Deleting the surveillance of a 
nonsafety-related equipment protection function from TS likewise 
does not change the probability or consequences of analyzed accident 
scenarios. Dose consequences remain unchanged by this request.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    A possibility for an accident or malfunction of a different type 
than any evaluated previously in SQN's FSAR [Final Safety Analysis 
Report] is not created; nor is the possibility for an accident or 
malfunction of a different type. The proposal does not alter the way 
any structure, system or component functions and does not modify the 
manner in which the plant is operated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The margin of safety has not been reduced since the test 
methodologies are not being changed and LCO [Limiting Condition for 
Operation] allowed outage times are not being changed. Deleting the 
surveillance of a nonsafety-related equipment protection function 
from TS likewise does not reduce the margin of safety. The results 
of accident analysis remain unchanged by this request.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

The Cleveland Electric Illuminating Company, Centerior Service Company, 
Duquesne Light Company, Ohio Edison Company, Pennsylvania Power 
Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power 
Plant, Unit 1, Lake County, Ohio

    Date of amendment request: October 27, 1998.
    Description of amendment request: The proposed amendment would 
modify the existing Minimum Critical Power Ratio (MCPR) Safety Limit 
contained in Technical Specification 2.1.1.2. The change would apply 
additional conservatism by modifying the MCPR Safety Limit values, as 
calculated by General Electric, by maintaining the limit of 1.09 for 
two recirculation loop operation and by increasing the limit from 1.10 
to 1.11 for single loop operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    There is no change to any plant equipment. Per USAR Section 
4.2.1, the fuel system design bases are provided in General Electric 
Standard Application for Reactor Fuel (GESTAR II). The Minimum 
Critical Power Ratio (MCPR) Safety Limit protects the fuel in 
accordance with the design basis. The MCPR Safety Limit calculations 
limit the bundle power to ensure the critical power ratio remains 
unchanged. Therefore, there is not an increase in the probability of 
transition boiling. The basis of the MCPR Safety Limit calculation 
remains the same,

[[Page 66604]]

ensuring that greater than 99.9% of all fuel rods in the core avoid 
transition boiling if the limit is not violated. Therefore, there is 
no increase in the probability of the occurrence of a previously 
analyzed accident.
    The fundamental sequences of accidents and transients have not 
been altered. The MCPR Operating Limits are selected such that 
potentially limiting plant transients and accidents prevent the MCPR 
from decreasing below the MCPR Safety Limit anytime during the 
transient. Therefore, there is no impact on any of the limiting USAR 
Appendix 15B transients. The radiological consequences are the same 
as previously stated in the USAR, and as approved in the NRC Safety 
Evaluation for GESTAR II. Therefore, the consequences of an accident 
do not increase over previous evaluations in the USAR.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The MCPR Safety Limit values are designed to ensure that fuel 
damage from transition boiling does not occur in at least 99.9% of 
the fuel rods in the core as a result of the limiting postulated 
accident. The values are calculated in accordance with GESTAR II and 
the fuel vendor's interim implementing procedures, which incorporate 
cycle-specific parameters.
    The GESTAR II analysis has been accepted by the NRC as 
comprehensive for ensuring that fuel designs will perform within 
acceptable bounds. The MCPR Safety Limit ensures that the fuel is 
protected in accordance with the design basis. The function, 
location, operation, and handling of the fuel remain unchanged. In 
addition, the initiating sequence of events has not changed. 
Therefore, no new or different kind of accident is created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The MCPR Safety Limit values do not alter the design or function 
of any plant system, including the fuel. The new MCPR Safety Limit 
values were calculated using NRC-approved methods described in 
GESTAR II and the fuel vendor's interim implementing procedures, 
which incorporate cycle-specific parameters. The MCPR Safety Limit 
values are consistent with GESTAR II, the NRC Safety Evaluation of 
GESTAR II, the NRC Safety Evaluation Report for the Perry Nuclear 
Power Plant and its Supplements for USAR Sections 4.4.1 and 
15.0.3.3.1, and the Technical Specification Bases (Section 2.1.1.2) 
for the MCPR Safety Limit. This change incorporates a cycle-specific 
MCPR Safety Limit, as opposed to relying on the generic limit. 
Therefore, the implementation of the proposed change to the MCPR 
Safety Limit does not involve a reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Stuart A. Richards.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application request: October 27, 1998 (supersedes the April 
12, 1996, amendment request). This notice supersedes the staff's 
proposed no significant hazards consideration determination evaluation 
for the requested changes that was published on May 8, 1996 (61 FR 
20858).
    Description of amendment request: The proposed amendment 
application would change the technical specifications (TS) for the 
reactor coolant system and associated Bases to allow the installation 
of electrosleeves in the Callaway steam generators for two fuel cycles.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The electrosleeve configuration has been designed and analyzed 
in accordance with the requirements of the ASME [American Society of 
Mechanical Engineers] Code. The applied stresses and fatigue usage 
for the sleeve are bounded by the limits established in the ASME 
Code. ASME Code minimum material property values are used for the 
structural and plugging limit analysis. Mechanical testing has shown 
that the structural strength of nickel electrosleeves under normal, 
upset and faulted conditions provides margin to the acceptance 
limits. These acceptance limits bound the most limiting (3 times 
normal operating pressure differential) burst margin recommended by 
RG [Regulatory Guide] 1.121. Leakage testing for \5/8\'', \7/8\'', 
\11/16\'' and \3/4\'' tube sleeves has demonstrated that no 
unacceptable levels of primary to secondary leakage are expected 
during any plant condition.
    The sleeve nominal wall thickness (used for developing the 
depth-based plugging limit for the sleeve) is determined using the 
guidance of Regulatory Guide 1.121 and the pressure stress equation 
of Section III of the ASME Code. The limiting requirement of 
Regulatory Guide 1.121, which applies to part throughwall 
degradation, is that the minimum acceptable wall must maintain a 
factor of safety of three against tube failure under normal 
operating (design) conditions. A bounding set of design and 
transient loading input conditions was used for the minimum wall 
thickness evaluation in the generic evaluation. Evaluation of the 
minimum acceptable wall thickness for normal, upset and postulated 
accident condition loading per the ASME Code indicates these 
conditions are bounded by the design condition requirement minimum 
wall thickness.
    A bounding tube wall degradation growth rate per cycle and a NDE 
[Non-Destructive Examination] uncertainty has been assumed for 
determining the sleeve TS plugging limit. The sleeve wall 
degradation extent is determined by NDE. The degradation which would 
require plugging sleeved tubes is developed using the guidance of RG 
1.121 and is defined in BAW-10219P, to be 20% throughwall for any 
service induced degradation.
    The consequences of failure of the sleeve are bounded by the 
current steam generator tube rupture analysis included in the 
Callaway FSAR [Final Safety Analysis Report]. 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 (depending on the break 
location), and therefore, would result in lower total primary fluid 
mass release to the secondary system.
    A risk assessment for installation of Electrosleeves at Callaway 
Plant was performed for a two-cycle operating period. The results of 
this evaluation determined that sufficient margins against 
postulated tube rupture during bounding accident conditions exist 
for all types of degradation of the Electrosleeve material. The 
calculated probability of burst for a hypothetical population of 
10,000 axial flaws, 100% throughwall of the parent tube and 0.40'' 
long, is 4.4 x 10-11 at the end of the second operating cycle. The 
probability of burst for postulated circumferential flaws and pits 
is determined to be essentially zero.
    The proposed change does not adversely impact any other 
previously evaluated design basis accident or the results of LOCA 
[Loss of Coolant Accident] and non-LOCA accident analyses for the 
current technical specification minimum reactor coolant system flow 
rate. The results of the analyses and testing demonstrate that the 
electrosleeve is an acceptable means of maintaining tube integrity. 
Furthermore, per Regulatory Guide 1.83 recommendations, the sleeved 
tube can be monitored through periodic inspections with present NDE 
techniques. These measures demonstrate that installation of sleeves 
spanning degraded areas of the tube will restore the tube to a 
condition consistent with its original design basis.
    Conformance of the electrosleeve design with the applicable 
sections of the ASME Code and results of the leakage and mechanical 
tests, support the conclusion that installation of electrosleeves 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

[[Page 66605]]

    Electrosleeving does not represent a potential to adversely 
affect any plant component. Stress and fatigue analysis of the 
repair has shown that the ASME Code and Regulatory Guide 1.121 
criteria are not exceeded. Implementation of electrosleeving 
maintains overall tube bundle structural and leakage integrity at a 
level consistent to that of the originally supplied tubing during 
all plant conditions. Leak and mechanical testing of electrosleeves 
support the conclusions of the calculations that each sleeve retains 
both structural and leakage integrity during all conditions. 
Sleeving of tubes does not provide a mechanism resulting in an 
accident outside of the area affected by the sleeves. Any accident 
as a result of potential tube or sleeve degradation in the repaired 
portion of the tube is bounded by the existing tube rupture accident 
analysis.
    Implementation of sleeving will reduce the potential for primary 
to secondary leakage during a postulated steam line break while not 
significantly impacting available primary coolant flow area in the 
event of a LOCA. By effectively isolating degraded areas of the tube 
through repair, the potential for steam line break leakage is 
reduced. These degraded intersections now are returned to a 
condition consistent with the Design Basis. While the installation 
of a sleeve reduces primary coolant flow, the reduction is far below 
that caused by plugging. Therefore, far greater primary coolant flow 
area is maintained through sleeving versus plugging.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The electrosleeve repair of degraded steam generator tubes has 
been shown by analysis to restore the integrity of the tube bundle 
consistent with its original design basis condition, i.e., tube/
sleeve operational and faulted condition stresses are bounded by the 
ASME Code requirements and the repaired tubes are leaktight. The 
safety factors used in the design of sleeves for the repair of 
degraded tubes are consistent with the safety factors in the ASME 
Code used in steam generator design. The portions of the installed 
sleeve assembly which represent the reactor coolant pressure 
boundary can be monitored for the initiation and progression of 
sleeve/tube wall degradation, thus satisfying the requirements of 
Regulatory Guide 1.83. The portion of the tube bridged by the sleeve 
is effectively removed from the pressure boundary, and the sleeve 
then forms the new pressure boundary. The areas of the sleeved tube 
assembly which require inspection are defined in BAW-10219P.
    In addition, since the installed sleeve represents a portion of 
the pressure boundary, a baseline inspection of these areas is 
required prior to operation with sleeves installed. The effect of 
sleeving on the design transients and accident analyses has been 
reviewed based on the installation of sleeves up to the level of 
steam generator tube plugging coincident with the minimum reactor 
flow rate and the Callaway Safety Analysis.
    Provisional requirements cited in other NRC Safety Evaluation 
Reports addressing the implementation of sleeving have required the 
reduction of the individual steam generator normal operation primary 
to secondary leakage limit from 500 to 150 gpd [gallons per day]. 
Consistent with these evaluations, Union Electric will reduce the 
per steam generator leak rate of 500 gpd in TS 3.4.6.2.c to 150 gpd. 
The establishment of this leakage limit at 150 gpd provides 
additional safety margin. [The staff notes that this leakage limit 
has been incorporated into the Callaway Technical Specifications via 
license amendment #119 dated October 1, 1996.]
    Finally, Union Electric will reduce the tube plugging limit from 
48% through wall to 40% through wall to be consistent with NUREG-
1431. The establishment of the plugging limit at 40% through wall 
provides additional safety margin. [The staff notes that this 
plugging limit has been incorporated into the Callaway Technical 
Specifications via license amendment #119 dated October 1, 1996.]

    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: University of Missouri-
Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Project Director: William H. Bateman.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont 
Yankee Nuclear Power Station, Vernon, Vermont

    Date of amendment request: November 3, 1998.
    Description of amendment request: The licensee proposes to make 
administrative changes to the Technical Specifications to correct 
errors, add consistency within the Technical Specifications, and make 
nomenclature changes to support and enhance usability of the Technical 
Specifications.
    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, because:
    The proposed changes are purely administrative in nature and 
have no effect on plant hardware, plant design, safety limit 
setting, or plant system operation and therefore do not modify or 
add any initiating parameters that would significantly increase the 
probability or consequences of an accident previously evaluated.
    No new modes of operation are introduced by the proposed changes 
such that adverse consequences would result. Accordingly, the 
consequences of previously analyzed accidents are not affected by 
this proposed license amendment.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated, because:
    These changes do not affect the operation of any systems or 
components, nor do they involve any potential initiating events that 
would create any new or different kind of accident. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated for the 
Vermont Yankee Nuclear Power Station.
    3. Involve a significant reduction in a margin of safety, 
because:
    These proposed changes do not affect any equipment involved in 
potential initiating events or safety limits. Therefore, it is 
concluded that the proposed changes do not involve a significant 
reduction in a margin of safety.
    Administrative changes, as such, do not constitute any 
significant hazards considerations.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Project Director: Cecil O. Thomas.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: November 10, 1998.
    Description of amendment request: The proposed changes to North 
Anna Power Station (NAPS), Units 1 and 2, Technical Specification (TS) 
3.4.4 will clarify the operability requirements for the pressurizer 
heaters and eliminate a potential verbatim compliance issue associated 
with the pressurizer heaters and emergency power supply. The verbatim 
compliance issue was created when the Emergency Diesel Generator 
allowed outage time was changed from 72 hours to 14 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 66606]]

licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    Virginia Electric and Power Company has reviewed the 
requirements of 10 CFR 50.92 as they relate to the proposed changes 
for the North Anna Units 1 and 2 and determined that a significant 
hazards consideration is not involved. The proposed changes will 
revise the LCO [limiting condition for operation] 3.4.4 to require 
that the pressurizer have two groups of pressurizer heaters operable 
with a capacity of greater than or equal to 125 kW and capable of 
being powered from its associated emergency bus. The Action 
Statement will also be revised to focus on heater operability. The 
following is provided to support this conclusion.
    (a) Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The pressurizer heaters are not an initiator of any accident 
previously evaluated. As a result, the probability of any accident 
previously evaluated is not increased. The pressurizer heaters 
remain operable as assumed in the accident analysis to mitigate the 
consequences of any accident. Therefore, the proposed changes to 
clarify the operability requirements do not significantly increase 
the probability of occurrence or the consequences of any previously 
analyzed accident.
    (b) Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed Technical Specifications changes do not involve any 
physical alteration of the plant or changes in methods governing 
normal plant operation. Operation of and the design of the 
pressurizer heaters and the associated power supplies are not 
changed by the proposed changes. The proposed changes do not impose 
any new or eliminate any existing requirements. Therefore, it is 
concluded that no new or different kind of accident or malfunction 
from any previously evaluated has been created.
    (c) Does the change involve a significant reduction in a margin 
of safety?
    The proposed Technical Specifications changes will not reduce 
the margin of safety since the change has no effect on any safety 
analyses assumptions. The pressurizer heaters remain operable as 
assumed in the safety analysis to mitigate the consequences of any 
accident previously analyzed. The proposed changes only clarify the 
operability requirements for the pressurizer heaters and associated 
emergency power supplies. Therefore, the proposed changes do not 
result in a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams, 
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
23219.
    NRC Project Director: Herbert N. Berkow.

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.

Florida Power and Light Company, et al., Docket Nos. 50-335, and 50-
389, St. Lucie Plant, Unit Nos. 1, and 2, St. Lucie County, Florida

    Date of amendment request: October 29, 1998.
    Description of amendment request: Technical Specification changes 
(TS) relating to the implementation and automatic removal of certain 
reactor protection system trip bypasses to ensure that the meaning of 
explicit terms used in the TSs are consistent with the intent of the 
stated requirements.
    Date of publication of individual notice in the Federal Register: 
November 5, 1998 (63 FR 59809).
    Expiration date of individual notice: November 19, 1998.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see: (1) The 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

Baltimore Gas and Electric Company, Docket No. 50-318, Calvert Cliffs 
Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

    Date of application for amendment: July 20, 1998.
    Brief description of amendment: The amendment implements a 
modification that constitutes an unreviewed safety question as 
described in 10 CFR 50.59. The modification involves replacing the 
service water heat exchangers with new plate and frame heat exchangers 
having an increased thermal performance capability. The planned 
modification is similar to the one completed on Unit 1. In addition, by 
a separate letter dated July 20, 1998, the licensee submitted a request 
to obtain approval for a temporary one time cooling lineup needed to 
support emergency diesel generator operability for the installation of 
the Unit 2 service water heat exchanger replacement, which is currently 
being reviewed by the NRC

[[Page 66607]]

staff. Therefore, since the implementation of the proposed service 
water heat exchanger modification is dependent on the staff's issuance 
of the one time Technical Specification (TS) change regarding 
installation of the modification, this modification should not be 
implemented prior to the issuance of the one-time TS change for 
installing the modification.
    Date of issuance: November 5, 1998.
    Effective date: This license amendment is effective as of the date 
of its issuance to be implemented after the staff's issuance of the 
one-time TS change regarding the installation of the service water heat 
exchanger modification.
    Amendment No.: 203.
    Facility Operating License No. DPR-69: Amendment revised the 
Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: August 12, 1998 (63 FR 
43201).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 5, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of application for amendment: June 26, 1998.
    Brief description of amendment: The amendment modifies various 
Technical Specification pages to correct typographical errors, remove 
inadvertent replication of information, and updates various Bases 
sections.
    Date of issuance: November 10, 1998.
    Effective date: November 10, 1998.
    Amendment No: 178.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 23, 1998 (63 
FR 50933).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 10, 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, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: March 6, 1998, as supplemented 
September 11, 1998. The September 11, 1998, supplemental letter 
contained clarifying information only, and did not change the no 
significant hazards consideration determination.
    Brief description of amendment: The amendment revises Technical 
Specification 3.9.2 relating to the use of Post-Accident Monitoring 
Source Range neutron flux detectors as a compensatory measure in the 
event that one of the two required BF3 neutron flux detectors becomes 
inoperable during Mode 6 operations (refueling).
    Date of issuance: November 12, 1998.
    Effective date: November 12, 1998.
    Amendment No: 180.
    Facility Operating License No. DPR-23: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 3, 1998 (63 FR 
30262).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 12, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application of amendments: September 17, 1998, as 
supplemented October 15, 1998.
    Brief description of amendments: The amendments revised the Updated 
Final Safety Analysis Report to perform a Keowee Emergency Power 
Engineered Safeguards Functional Test during the 1998 Unit 3 refueling 
outage at Oconee.
    Date of Issuance: November 12, 1998.
    Effective date: As of the date of issuance to be implemented during 
the 1998 Unit 3 refueling outage.
    Amendment Nos.: Unit 1--233; Unit 2--233; Unit 3--232.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: September 30, 1998 (63 
FR 52304).
    The October 15, 1998, letter provided clarifying information that 
did not change the scope of the September 17, 1998, application and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 12, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.

GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear 
Station, Dauphin County, Pennsylvania

    Date of application for amendment: December 2, 1996.
    Brief description of amendment: This amendment would revise audit 
frequency requirements and relocate them from the Technical 
Specifications to the Quality Assurance Plan.
    Date of issuance: November 12, 1998.
    Effective date: This amendment is effective immediately to be 
implemented written 60 days.
    Amendment No.: 52.
    Facility Operating License No. DPR-73: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 (62 FR 
40850).
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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 11, 1998, as supplemented 
by letter dated October 9, 1998.
    Brief description of amendments: The amendments modify the 
technical specifications (TS) for San Onofre Nuclear Generating Station 
Unit Nos. 2 and 3 to implement 10 CFR Part 50 Appendix J, Option B for 
performance-based reactor containment leakage testing.
    Date of issuance: November 6, 1998.
    Effective date: November 6, 1998, to be implemented within 30 days 
from the date of issuance.
    Amendment Nos.: Unit 2 -144; Unit 3 -135.
    Facility Operating License No. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.

[[Page 66608]]

    Date of initial notice in Federal Register: September 9, 1998 (63 
FR 48265).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 6, 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.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: July 6, 1998.
    Brief description of amendments: Relocates the description of the 
reactor coolant system design features in Technical Specification 5.4 
to the Updated Final Safety Analysis Report, which already contains the 
information.
    Date of issuance: November 18, 1998.
    Effective date: November 18, 1998, to be implemented within 30 
days.
    Amendment Nos.: Unit 1--Amendment No. 98; Unit 2--Amendment No. 85.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 9, 1998 (63 
FR 48266).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 18, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas.

    Date of amendment request: July 6, 1998 , as supplemented on 
October 28, 1998.
    Brief description of amendments: Relocate the Technical 
Specification 3/4.3.3.3 requirements for Seismic Instrumentation to the 
Technical Requirements Manual.
    Date of issuance: November 18, 1998.
    Effective date: November 18, 1998, to be implemented within 30 
days.
    Amendment Nos.: Unit 1--Amendment No. 99; Unit 2--Amendment No. 86.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 9, 1998 (63 
FR 48267).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 18, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas.

    Date of amendment request: July 6, 1998, as supplemented on October 
28, 1998.
    Brief description of amendments: Relocates the Technical 
Specification 3/4.7.13 requirements for the Area Temperature Monitoring 
System to the Technical Requirements Manual.
    Date of issuance: November 18, 1998.
    Effective date: November 18, 1998, to be implemented within 30 
days.
    Amendment Nos.: Unit 1--Amendment No. 100; Unit 2--Amendment No. 
87.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendment 
revises the Technical Specifications.
    Date of initial notice in Federal Register: September 9, 1998 (63 
FR 48267). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 18, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee.

    Date of application for amendments: February 13, 1998 (TS 97-07).
    Brief description of amendments: The amendments incorporate new 
main steam isolation valve (MSIV) requirements that are consistent with 
NUREG-1431, the Westinghouse Standard Technical Specifications (TS), 
including testing requirements for the MSIVs that ensure the valves 
close on an automatic actuation signal.
    Date of issuance: November 17, 1998.
    Effective date: As of the date of issuance to be implemented no 
later than 45 days after issuance.
    Amendment Nos.: 236 and 226.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19980).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 17, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: June 26, 1998 (TS 98-02).
    Brief description of amendments: The amendments change the 
Technical Specifications and their Bases to lower the specific activity 
limit for the primary coolant system from 1.0 microcurie/gram dose 
equivalent iodine-131 to 0.35 microcurie/gram, as provided for in NRC 
Generic Letter 95-05, ``Voltage-Based Repair Criteria for Westinghouse 
Steam Generator Tubes Affected by Outside Diameter Stress Corrosion 
Cracking.'' This change allows a proportional increase in main steam 
line break induced primary-to-secondary leakage when implementing the 
alternate steam generator tube repair criteria, which the NRC has 
already approved for Sequoyah Nuclear Plant, Units 1 and 2.
    Date of issuance: November 17, 1998.
    Effective date: As of the date of issuance to be implemented no 
later than 45 days after issuance.
    Amendment Nos.: 237 and 227.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: July 15, 1998 (63 FR 
38205).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 17, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

[[Page 66609]]

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, 
Unit 1, (WBN) Rhea County, Tennessee

    Date of application for amendment: August 5, 1998 (TS 98-008).
    Brief description of amendment: This amendment is in response to 
your application dated August 5, 1998. The amendment revises the WBN 
Technical Specifications (TS) and associated TS Bases to allow up to 4 
hours to make the residual heat removal suction relief valve available 
as a cold overpressure mitigation system relief path.
    Date of issuance: November 10, 1998.
    Effective date: November 10, 1998.
    Amendment No.: 14.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 23, 1998 (63 
FR 50940).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 10, 1998.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402.

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: July 10, 1996 (TXX-96405), as 
supplemented by letters dated October 1, 1996 (TXX-96475), and July 1, 
1998 (TXX-98159).
    Brief description of amendments: The amendment would take credit 
for the addition of train oriented Fan Coil Units for each UPS and 
Distribution Room and would provide redundancy to the existing Air 
Conditioning (A/C) Units (TS 3/4.7.11 and its associated bases).
    Date of Issuance: Date of issuance: November 18, 1998.
    Effective date: November 18, 1998, to be implemented within 30 
days.
    Amendment Nos.: Unit 1--Amendment No. 61; Unit 2--Amendment No. 47.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 12, 1997 (62 
FR 6579).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 18, 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.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: May 7, 1998.
    Brief description of amendment: This amendment revises Technical 
Specification 5.4, ``Fuel Storage,'' to increase the allowable mass of 
uranium-235 (U235) per axial centimeter for fuel storage. 
The requested change will allow the use of new Siemens Power 
Corporation heavy fuel assembly designs.
    Date of Issuance: November 12, 1998.
    Effective date: November 12, 1998.
    Amendment No.: 141.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 17, 1998 (63 FR 
33111).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 12, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.

    Dated at Rockville, Maryland, this 24th day of November 1998.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director Division of Reactor Projects--III/IV Office of Nuclear 
Reactor Regulation.
[FR Doc. 98-31931 Filed 12-1-98; 8:45 am]
BILLING CODE 7590-01-P