[Federal Register Volume 64, Number 86 (Wednesday, May 5, 1999)]
[Notices]
[Pages 24192-24211]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-11119]


-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from April 10, 1999, through April 23, 1999. The 
last biweekly notice was published on April 21, 1999 (64 FR 19554).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed no Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
Involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW, Washington, DC. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By June 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

[[Page 24193]]

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.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: April 12, 1999.
    Description of amendment request: The proposed one-time technical 
specification (TS) change, effective through September 30, 1999, 
provides a Required Action and Completion Time for the Ultimate Heat 
Sink (UHS) in the event that service water temperature exceeds the 
current 95 deg.F surveillance limit. It involves an allowance to 
continue operation for a period of 8 hours with the UHS at a 
temperature greater than the temperature limits provided in TS Limiting 
Condition of Operation 3.7.8, ``Ultimate Heat Sink (UHS)'' and provides 
an upper UHS temperature limit beyond which plant shutdown is required.
    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:

    Carolina Power & Light (CP&L) Company has evaluated the proposed 
Technical Specification change and has concluded that it does not 
involve a significant hazards consideration. The conclusion is in 
accordance with the criteria set forth in 10 CFR 50.92. The bases 
for the conclusion that the proposed change does not involve a 
significant hazards consideration are discussed below.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change does not involve any physical alteration of 
plant systems, structures or components. The proposed change allows 
plant operation for a short period of time when the service water 
temperature exceeds 95 deg.F, requires an hourly surveillance when 
service water temperature exceeds 95 deg.F, provides an upper UHS 
temperature limit beyond which a plant shutdown is required, and 
specifies an expiration date beyond which the current requirements 
are restored. If the service water temperature is restored within 
the allowed time, a plant shutdown is not required. This minimizes 
plant transients, which reduces the probability of a reactor trip 
and the resulting challenges to mitigating systems. A service water 
temperature of up to 99 deg.F does not increase the failure rate of 
systems, structures or components because the systems, structures, 
and components are designed for higher temperatures than at which 
they operate.
    The Service Water (SW) System temperature is not assumed to be 
an initiating condition of any accident evaluated in the safety 
analysis report. Therefore, the allowance of a limited time for 
service water temperature to be in excess of 95 deg.F does not 
involve an increase in the probability of an accident previously 
evaluated in the safety analysis report (SAR). The SW System 
supports operability of safety related systems used to mitigate the 
consequences of an accident. The service water temperature is not 
expected to increase significantly beyond 95 deg.F due to the 
limited time allowed by the proposed change in conjunction with the 
generally slow rate of temperature increase experienced from thermal 
changes in Lake Robinson. The capability of components to perform 
their safety related function is not affected up to a service water 
temperature of 99 deg.F with the exception of the Containment Air 
Recirculation Fan Coolers. The heat removal capacity of the 
Containment Air Recirculation Fan Coolers is not expected to be 
significantly reduced by a small increase in service water 
temperature. If heat removal is not significantly reduced, 
containment pressure and leakage will not be significantly 
increased, and the doses from containment leakage will not be 
significantly increased. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated in the SAR.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve any physical alteration of 
plant systems, structures or components. A service water temperature 
of up to 99 deg.F does not introduce new failure mechanisms of 
systems, structures or components not already considered in the SAR 
because the systems, structures, and components are designed for 
higher temperatures than at which they operate. Therefore, the 
possibility of a new or different kind of accident from any accident 
previously evaluated is not created.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change will allow a small increase in service water 
temperature above the design basis limit for the SW System and delay 
by 8 hours the requirement to shutdown the plant when the service 
water system design limit is exceeded. There are design margins 
associated with systems, structures and components that are cooled 
by the service water system that are affected. The capability of 
components to perform their safety related function is not affected 
up to a service water temperature [of] 99 deg.F with the exception 
of the Containment Air Recirculation Fan Coolers. The Containment 
Air Recirculation Fan Coolers remove heat from containment to 
mitigate containment pressure and temperature following a MSLB [main 
steamline break] inside containment or a Large Break LOCA [loss-of-
coolant accident] inside containment. An increase in service water 
temperature in excess of the design limit due to hot weather 
conditions is expected to be small due to the limited time allowed 
by the proposed change in conjunction with the generally slow rate 
of temperature increase experienced from thermal changes in Lake 
Robinson. Therefore, the effect on the Containment Air

[[Page 24194]]

Recirculation Fan Coolers' heat removal capacity and the resulting 
containment pressure and temperature is expected to be small. 
Therefore, there is no significant reduction in margin of safety 
associated with this change.

    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: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Sheri R. Peterson.

Commonwealth Edison Company, Docket No. 50-254, Quad Cities Nuclear 
Power Station, Units 1, Rock Island County, Illinois

    Date of amendment request: March 30, 1999.
    Description of amendment request: The amendment would revise the 
Quad Cities Nuclear Power Station, Unit 1 Technical Specifications (TS) 
by changing the Surveillance Requirements (SR) 4.6.E.2 to allow a one-
time extension of the 18-month requirement to pressure set test or 
replace one half of the Main Steam Safety Valves (MSSVs) to an interval 
of 24 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes request a one-time change to the 
surveillance requirement for the MSSVs and Target Rock S/RV [Safety 
Relief Valve]. The surveillance interval between MSSVs and Target 
Rock S/RV testing is not a precursor assumed in any previously 
analyzed accident. Therefore, the probability of a previously 
evaluated accident has not been increased.
    The proposed extension is consistent with the ASME Code 
requirement to test 20% of the sample population every 24 months 
with all of the valves in the sample group being tested every 60 
months. The proposed changes are also consistent with NUREG 1433, 
Revision 1, and do not adversely affect existing plant safety 
margins or the reliability of the equipment assumed to operate in 
the safety analysis. Operating experience and excellent materiel 
condition of the MSSVs and Target Rock S/RV support the expectation 
that they will continue to perform their intended function. 
Therefore, the consequences of a previously evaluated accident have 
not been increased.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No new equipment is required, nor will the MSSVs and Target Rock 
S/RV be operated in a different manner during the period of the 
extended surveillance interval. The proposed changes are consistent 
with NUREG 1433, Revision 1, requirements for safety valve 
surveillance intervals as well as the ASME Code requirements for 
testing safety valves. Operating experience and superior materiel 
condition of the MSSVs and Target Rock S/RV support the expectation 
that they will continue to perform their intended function. 
Therefore, the possibility of a new or different accident has not 
been increased.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed amendment represents an extension to the current TS 
SRs that would otherwise be provided generically by the ASME Code. 
The proposed changes are also consistent with NUREG-1433, Revision 
1, and do not adversely affect existing plant safety margins or the 
reliability of the equipment assumed to operate in the safety 
analysis. The proposed changes have been evaluated and found to be 
acceptable for use at Quad Cities Nuclear Power Station based on 
system safety analysis requirements and operational performance. The 
MSSVs and Target Rock S/RV provisions continue to be adequately 
maintained during plant operation. The proposed changes to the MSSVs 
and Target Rock S/RV surveillance interval do not significantly 
reduce existing plant safety margins since excellent materiel 
condition and acceptable surveillance test results support the 
expectation that no significant degradation will occur over the 
extended interval.
    The proposed changes are based on NRC accepted provisions at 
other operating plants that are applicable at Quad Cities Nuclear 
Power Station and maintain necessary levels of system or component 
reliability.
    The proposed amendment for Quad Cities Nuclear Power Station 
will not reduce the availability of systems required to mitigate 
accident conditions.
    Therefore, these changes do not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities 
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

    Date of amendment request: March 30, 1999.
    Description of amendment request: This amendment request proposes 
to change the Technical Specifications (TSs) to allow an alternate 
methodology for quantifying Reactor Coolant System (RCS) leakage when 
the normal RCS leakage detection system is inoperable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The current Technical Specifications require a periodic 
measurement of RCS leakage. The normal method for quantifying RCS 
leakage is to use the DWFDS [Drywell Floor Drain Sump] and DWEDS 
[Drywell Equipment Drain Sump] flow totalizers. The proposed TS 
change would allow an alternate method for quantifying RCS leakage 
when a flow totalizer is not available. The proposed change has no 
impact on the frequency for monitoring RCS leakage and would only be 
used for a maximum of 30 days while the normal leakage monitoring 
system is being restored to an operable condition. The alternate 
methodology for quantifying leakage has a measurement sensitivity 
that is consistent with the normal method. The proposed change does 
not impact any system structure or component used to mitigate the 
consequences of an accident and there will be no change in the types 
or significant increase in the amounts of any effluents released 
offsite.
    Therefore this proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change involves no physical modifications to any 
system, structure or component used to mitigate the consequences of 
an accident. The operation of the DWEDS and DWFDS are not being 
altered in any way that could affect their ability to function 
during an accident condition.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The current TS's require a periodic measurement of RCS leakage. 
The normal method for quantifying RCS leakage is to use

[[Page 24195]]

the DWFDS and DWEDS flow totalizers. The proposed technical 
specifications change would allow an alternate method for 
quantifying RCS leakage when a flow totalizer is inoperable. The 
proposed change has no impact on the frequency for monitoring RCS 
leakage and would only be used for a maximum of 30-days while the 
normal leakage monitoring system is being restored to an operable 
condition. The proposed alternate methodology for quantifying 
leakage has a measurement sensitivity that is consistent with the 
normal method.
    Therefore, these changes do not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities 
Nuclear Power Station, Units and, Rock Island County, Illinois

    Date of amendment request: March 30, 1999.
    Description of amendment request: This amendment request proposes 
to revise license conditions in each of the respective Operating 
Licenses to delete those license conditions that no longer apply, make 
an editorial change in the Unit 1 license, and provide clarifying 
information regarding the license condition concerning equalizer valve 
restrictions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 FR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The initial conditions and methodologies used in the accident 
analyses remain unchanged. The proposed changes do not change or 
alter the design assumptions for the systems or components used to 
mitigate the consequences of an accident. Therefore, accident 
analyses results are not impacted.
    The proposed changes delete various license conditions that have 
been completed, make editorial changes, and provide clarifying 
information. The changes are administrative. No physical or 
operational changes to the facility will result from the proposed 
changes.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not affect the design or operation of 
any system, structure, or component in the plant. The safety 
functions of the related structures, systems, or components are not 
changed in any manner, nor is the reliability of any structures , 
systems, or component reduced. The changes do not affect the manner 
by which the facility is operated and do not change any facility 
design feature, structure, system, or component. No new or different 
type of equipment will be installed.
    The proposed changes delete various license conditions that have 
been completed, make editorial changes, and provide clarifying 
information. The changes are administrative. No physical or 
operational changes to the facility will result from the proposed 
changes.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    Does the change involve a significant reduction in the margin of 
safety for the following reasons:
    The proposed changes are administrative in nature and have no 
impact on the margin of safety of any Technical Specification. There 
is no impact on safety limits or limiting safety system settings. 
The changes do not affect any plant safety parameters or setpoints. 
The proposed changes delete various license conditions that have 
been completed, make editorial changes, and provide clarifying 
information. No physical or operational changes to the facility will 
result from the proposed changes.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: March 25, 1999.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TS) to redefine the ``trip setpoint'' in a 
number of locations as the ``nominal trip setpoint.'' The current 
definition results in upper-or lower-bound numerical values not to be 
exceeded for setpoints. This proposed new definition would permit the 
setpoints to be set within a tolerance range around the number 
specified in various tables. The TS locations affected are: Table 
3.3.1-1, ``Reactor Trip System Instrumentation;'' Table 3.3.2-1, 
``Engineered Safety Feature Actuation Instrumentation;'' Surveillance 
Requirement 3.3.5.2; Table 3.3.6-1, ``Containment Purge and Exhaust 
Isolation Instrumentation;'' and Limiting Condition of Operation (LCO) 
3.4.12. Sections of the associated TS Bases document would also be 
revised to reflect the TS changes.
    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?
    No. The proposed changes are consistent with the current 
licensing basis for Catawba Nuclear Station, the setpoint 
methodology used to develop the Trip Setpoints, the Catawba Safety 
Analyses, and current station calibration procedures and practices. 
The Reactor Trip System and Engineered Safety Features Actuation 
System are not accident initiating systems; they are accident 
mitigating systems. Therefore, these proposed changes will have no 
impact on any accident probabilities. Accident consequences will not 
be affected, as no changes are being made to the plant which will 
involve a reduction in reliability of these systems. Consequently, 
any previous evaluations associated with accidents will not be 
affected by these changes.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed changes are consistent with the current 
licensing basis for Catawba Nuclear Station, the setpoint 
methodology used to develop the Trip Setpoints, the Catawba Safety 
Analyses, and current station calibration procedures and practices. 
No changes are being made to actual plant hardware which will result 
in any new accident causal mechanisms. Also, no changes are being 
made to the way in which the plant is being operated. Therefore, no

[[Page 24196]]

new accident causal mechanisms will be generated. Consequently, 
plant accident analyses will not be affected by these changes.
    3. Does this change involve a significant reduction in a margin 
of safety?
    No. The proposed changes are consistent with the current 
licensing basis for Catawba Nuclear Station, the setpoint 
methodology used to develop the Trip Setpoints, the Catawba Safety 
Analyses, and current station calibration procedures and practices. 
Margin of safety is related to the confidence in the ability of the 
fission product barriers to perform their design functions during 
and following accident conditions. These barriers include the fuel 
cladding, the reactor coolant system, and the containment system. 
The performance of these barriers will not be degraded by the 
proposed changes. Consequently, plant safety analyses will not be 
affected by these 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: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina.
    NRC Section Chief: Richard L. Emch, Jr.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of amendment request: April 9, 1999.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) to add Limiting Condition for 
Operation (LCO) 3.0.6 and its associated bases. This change would allow 
equipment that has been removed from service or declared inoperable in 
compliance with the TS Action statement to be returned to service under 
administrative controls solely to perform testing required to 
demonstrate its operability or the operability of other equipment. The 
proposed change is consistent with TS 3.0.5 as discussed in NUREG-1432, 
Revision 1, ``Standard Technical Specifications for Combustion 
Engineering Plants.'' TS 3.0.2 would also be modified to reflect that 
TS 3.0.6 is an exception to TS 3.0.2.
    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. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change would allow an orderly return to service of 
inoperable equipment. This change does not alter the functional 
characteristics of any plant component and does not allow any new modes 
of operation of any component. The accident mitigation features of the 
plant are not affected by the proposed amendment request. Therefore, 
this proposed amendment would not result in a significant change in the 
types or significant increase in the amounts of any effluents that may 
be released off site. No modifications to the plant have been proposed 
due to this amendment request. The proposed change would permit 
equipment removed from service to comply with required actions to be 
returned to service under administrative controls to verify the 
operability of the equipment being returned to service or of other 
related equipment. Although returning inoperable equipment to service 
for testing may temporarily compromise single failure criteria, 
administrative controls will ensure the time involved will be limited 
to only that required to demonstrate component or system operability. 
This LCO provides an acceptable method of restoring equipment to 
service for the sole purpose of demonstrating its operability or the 
operability of other related equipment. Therefore, this change does not 
involve a significant increase in the probability or consequences of 
any accident previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    No modifications were made to the plant due to this amendment 
request. The proposed change does not alter the functional 
characteristics of any plant component and does not allow any new modes 
of operation for any component. This proposed amendment would 
facilitate the testing of equipment in its design configuration to 
demonstrate operability. The use of TS 3.0.6 would be limited to the 
time absolutely necessary to perform the test. Therefore, this 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 addition of TS 3.0.6 is considered necessary to establish an 
allowance that is not formally recognized in the current TSs. Without 
this allowance, situations can arise in which certain components could 
not be restored to operable status without requiring a plant shutdown. 
It is not the intent that the TSs preclude the return to service of a 
component to confirm its operability. This allowance is deemed to 
represent a more stable, safe operation than requiring a plant shutdown 
to complete the restoration and confirmatory testing. The time period 
during which the equipment is returned to service in conflict with the 
requirements of the TS Action statement is limited to the time 
absolutely necessary to perform the indicated surveillance requirement. 
TS 3.0.6 does not provide time to perform any other preventive or 
corrective maintenance. The period of time during which the equipment 
is returned to service will be limited by administrative controls and 
is considered very small. Therefore, the probability of an accident 
during that time period is also very small and is considered to be 
insignificant. Thus, it can be concluded that the proposed change does 
not affect the current margin of safety.
    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, NW, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: March 9, 1999.
    Description of amendment request: The proposed change would modify 
the Technical Specifications to increase the inservice inspection 
interval, and reduce the scope of volumetric and surface examinations 
for the reactor coolant pump flywheels.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:

    The Nuclear Regulatory Commission has provided standards in 10 
CFR 50.92(c) for

[[Page 24197]]

determining whether a significant hazard exists due to a proposed 
amendment to an Operating License for a facility. A proposed 
amendment involves no significant hazards consideration if operation 
of the facility in accordance with the proposed changes would: (1) 
Not involve a significant increase in the probability or 
consequences of an accident previously evaluated; (2) Not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated; or (3) Not involve a significant reduction in 
a margin of safety. The Davis-Besse Nuclear Power Station has 
reviewed the proposed changes and determined that a significant 
hazards consideration does not exist because operation of the Davis-
Besse Nuclear Power Station, (DBNPS) Unit No. 1, in accordance with 
these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators, 
conditions, or assumptions are affected by the proposed changes to 
Technical Specification Surveillance Requirement 4.4.10.1.a in the 
frequency and scope of volumetric and surface examinations for the 
Reactor Coolant Pump (RCP) motor flywheels.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because changes in the frequency and 
scope of volumetric and surface examinations for the RCP motor 
flywheels will not affect any previously evaluated accidents. 
Accidents associated with failure of the flywheel were not evaluated 
in the DBNPS Updated Safety Analysis Report (USAR). The design, 
fabrication, and testing of flywheels in accordance with the 
guidance found in NRC Regulatory Guide 1.14, ``Reactor Coolant Pump 
Flywheel Integrity,'' Revision 1, August 1975, minimizes the 
potential for flywheel failure. The proposed changes have been 
demonstrated to maintain conservative testing requirements for the 
flywheels.
    2. Not create the possibility of a new or different kind of 
accident from any previously evaluated because changes in the 
frequency and scope of volumetric and surface examinations for the 
RCP motor flywheels will not affect the reliability of RCP motor 
flywheels. No new failure mode is introduced since the proposed 
changes do not involve a modification or change in operation of any 
plant systems, structures, or components.
    3. Not involve a significant reduction in the margin of safety. 
As shown in Westinghouse Topical Report WCAP-14535A, ``Topical 
Report on Reactor Coolant Pump Flywheel Inspection Elimination,'' 
November 1996, RCP motor flywheels have been inspected for twenty 
years without any service induced flaws being identified. 
Additionally, the analyses demonstrated that the flywheels are 
manufactured from high quality steel, have a high fracture 
toughness, and have a very high flaw tolerance. The topical report 
indicates that the flywheels could be operated for forty years 
without inspection, and there would be no significant increase in 
the probability of failure of the flywheels. However, inspections 
are proposed to continue at a frequency of once every ten years as a 
conservative measure. Thus, the margin of safety is not reduced 
significantly by the proposed change in inspection frequency.
    Based on the above, the Davis-Besse Nuclear Power Station has 
determined that the License Amendment Request does not involve a 
significant hazards consideration. As this License Amendment Request 
concerns a proposed change to the Technical Specifications that must 
be reviewed by the Nuclear Regulatory Commission, this License 
Amendment Request does not constitute an unreviewed safety question.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Anthony J. Mendiola.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: March 31, 1999.
    Description of amendment request: The proposed change would modify 
Cooper Nuclear Station's technical specification administrative 
controls for unit staff qualifications for the shift supervisor, senior 
operator, licensed operator, shift technical advisor, and radiological 
manager.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed change provides enhancement to the current 
requirements and clarifies the qualifications and training 
requirements for the shift supervisor, senior operator, licensed 
operator, shift technical advisor, and Radiological Manager. This 
provides additional assurance that these personnel are properly 
trained and qualified for their positions; therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change will not create the possibility of a new or 
different kind of accident than evaluated in the Updated Safety 
Analysis Report (USAR). The proposed change provides enhancement to 
the current requirements and clarifies the qualifications and 
training requirements for the shift supervisor, senior operator, 
licensed operator, shift technical advisor, and Radiological 
Manager. The revised administrative controls for unit staff 
qualifications are an enhancement to the current requirements; 
therefore, the proposed change does not create the possibility of a 
new or different kind of accident.
    The proposed change will not create a significant reduction in 
the margin of safety. The proposed change provides enhancement to 
the current requirements and clarifies the qualifications and 
training requirements for the shift supervisor, senior operator, 
licensed operator, shift technical advisor, and Radiological 
Manager. This provides additional assurance that these personnel are 
properly trained and qualified for their positions; therefore, the 
proposed change will not create a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Auburn Memorial Library, 1810 
Courthouse Avenue, Auburn, NE 68305.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: March 31, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Table 3.6.1.2-1, ``Allowable Leak 
Rates through Valves in Potential Bypass Leakage Paths,'' by adding two 
relief valves, with associated leak rate criteria, to be installed on 
the drywell equipment drain line and drywell floor drain line during 
the refueling outage in the spring of 2000. Specifically:
    (i) For the drywell equipment drain line, the reference to the 
inboard isolation valve (2DER*MOV119) would be replaced with a 
reference to the isolation valve and its associated relief valve 
(2DER*MOV119 and 2DER*RV344);
    (ii) For the drywell floor drain line, the reference to the inboard 
isolation valve (2DFR*MOV121) would be

[[Page 24198]]

replaced with a reference to the isolation valve and its associated 
relief valve (2DFR*MOV121 and 2DFR*RV228); and
    (iii) A footnote for both above changes would be added to state, 
``For valves 2DER*MOV 119 and 2DER*RV344, and likewise for valves 
2DFR*MOV121 and 2DFR*RV228, this limit shall be the combined allowable 
leak rate and not the per valve allowable leak rate.''
    The two relief valves would be installed to protect the drain line 
penetrations against overpressure, consistent with Generic Letter 96-
06, ``Assurance of Equipment Operability and Containment Integrity 
During Design-Basis Accident Conditions.'' The allowable leak rates 
currently specified in TS Table 3.6.1.2-1 for the drywell equipment and 
drywell floor drain line penetrations will not be increased as a result 
of the hardware modifications or proposed TS amendment.
    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 [NMP2], in accordance 
with the proposed amendment, will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment will add one relief valve on the drywell 
equipment drain line (penetration 2DER*Z40) and one relief valve on 
the drywell floor drain line (penetration 2DFR*Z39). These valves 
will be installed on piping between the inboard containment 
isolation valve and the primary containment wall. These drain lines 
represent potential bypass leakage paths from the primary 
containment to the environment and are subject to maximum allowable 
isolation valve leak rates, as specified in Table 3.6.1.2-1 of the 
Technical Specifications (TS). The purpose of adding relief valves 
is to protect the piping between the inboard and outboard isolation 
valves against thermally induced overpressure under postulated 
accident conditions when both isolation valves close, and the fluid 
trapped between them may heat up and expand. The new relief valves 
and piping will not cause any existing plant design, operating, or 
testing limits to be exceeded. The relief valve installations will 
meet standards and specifications currently applicable to the 
penetrations being modified. The relief valve configuration, set 
pressure, and testing meet applicable NRC guidance. No different 
precursors or new accident initiators are introduced as the result 
of the proposed modification. Therefore, this proposed amendment 
does not involve a significant increase in the probability of an 
accident previously evaluated.
    The existing requirements relating to allowable bypass leakage 
for the two penetrations affected by this modification, will not be 
changed. No new bypass leakage paths to the environment will be 
created and no new failure modes will be introduced. Should the 
relief valves open and fail to close, the effectiveness of the 
containment and other fission product barriers will not be 
compromised. As a result, accident dose rates will remain unchanged 
and within the limits of 10 CFR 50, Appendix A, General Design 
Criterion 19, and 10 CFR 100. None of the accident assumptions 
described in Section 6.2, titled ``Containment Systems'' and Chapter 
15, titled ``Accident Analysis,'' of the NMP2 Updated Safety 
Analysis Report (USAR) is adversely affected by the proposed 
modifications. Therefore, this proposed amendment does not involve a 
significant increase in the 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 isolation valves associated with penetrations 2DER*Z40 
(drywell equipment drain line) and 2DFR*Z39 (drywell floor drain 
line) perform an accident mitigation function by isolating the 
containment during and after certain postulated accidents. The 
addition of relief valves between the inboard and outboard isolation 
valves will enhance the capability of the existing isolation valves 
to perform their function without the risk of failure due to piping 
overpressurization. Consistent with the guidance in Generic Letter 
96-06, the consequences of a stuck-open relief valve malfunction 
have been evaluated and are acceptable. Should the relief valve fail 
to close after opening, the existing outboard isolation valve will 
perform its function to isolate the containment. Therefore, 
operation of NMP2 in accordance with this 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 proposed installation of the relief valves will not 
adversely affect primary containment integrity, the maximum 
allowable leak rates for the affected penetrations, any other 
fission product barriers, or any plant safety/operational limits. 
The relief valves will assure that the associated isolation valves 
do not fail as the result of piping overpressure during and after 
postulated accidents, which will preserve the radiological margin of 
safety. Therefore, 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 Section Chief: S. Singh Bajwa

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423, 
Millstone Nuclear Power Station, Unit No. 3, New London County, 
Connecticut

    Date of amendment request: March 2, 1999.
    Description of amendment request: The proposed amendment would 
require two service water (SW) pumps and their associated strainers to 
be operable to declare a service water system (SWS) loop operable. The 
proposed amendment would also (1) modify the existing action statement 
to take into account one or more service water pump(s) or strainers 
being inoperable and (2) make changes to the appropriate Bases section.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    NNECO has reviewed the proposed revision in accordance with 10 
CFR50.92 and has concluded that the revision does not involve any 
Significant Hazards Considerations (SHC). The basis for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
satisfied. The proposed Technical Specification revision does not 
involve an SHC because the revision would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed TS [Technical Specification] change adds an 
additional AOT [allowed outage time] for one of four of the service 
water pumps/strainers in the SWS. The capabilities of the SWS were 
evaluated in order to ensure that a significant increase in the 
probability or consequences of the following previously evaluated 
accidents, LOP [loss of power], LOCA [loss-of-coolant accident] with 
concurrent LOP and secondary side piping break inside containment, 
are precluded by SWS mitigative functions. As the above DBA's 
[design basis accidents] are not caused by the failure of the SWS to 
operate, the SWS can not affect the probability of these accidents 
to occur.
    Since both pumps/strainers in each loop are covered by the 
ACTION statement in the TS when inoperable (due to failure or 
maintenance), and the proposed ACTION statement for two inoperable 
service water pumps in a single loop is consistent with the

[[Page 24199]]

current ACTION statement, there is no impact on the capability to 
maintain core decay heat removal following a DBA. Further, the 
revised TS will improve availability of the SWS. The LCO [limiting 
condition for operation] and ACTION statements help ensure that the 
SWS, including pumps/strainers, are kept in a condition which allows 
it to perform all its design functions including providing core 
decay heat removal and the SFP [spent fuel pool] cooling. As such, 
there is no affect on the consequences of previously evaluated 
accidents.
    Thus, it is concluded that the proposed revision does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The SWS is used to remove heat from the reactor plant auxiliary 
systems and other systems. Only one of four pumps is required to be 
operating during normal plant conditions. In addition, only one 100% 
capacity pump is required to provide the necessary flow to mitigate 
the consequences of a DBA. This change continues to require two 
pumps/strainers per loop to be operable and imposes strict controls 
on the AOT for the SWS pumps/strainers via the imposition of the LCO 
controls on the SWS. This assures that four service water pumps/
strainers will always be available or the plant will be in an ACTION 
STATEMENT. The SWS is used to mitigate the consequences of an 
accident and will not cause an accident.
    Thus, this proposed revision does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction [in] a margin of safety.
    This change will have no impact on the performance of any safety 
related system covered by the TS. This change explicitly defines the 
number of pumps/strainers required for the SWS to be considered 
OPERABLE and the ACTION required which specifies the AOT for 
inoperable components. The required flow rate for accident 
mitigation continues to be available to all ECCS [emergency core 
cooling system] components and their support systems. As such, this 
change does not increase the peak clad temperature for a DBA-LOCA.
    The proposed Technical Specification change adds an additional 
AOT for one of four of the service water pumps/strainers in the SWS. 
Two service water pumps/strainers are required to perform the design 
function of the SWS; one pump to mitigate the DBA and the other to 
reduce the potential of the SFP boiling which could occur if a 
service water pump is unavailable for SFP cooling after a design 
basis LOCA.
    The existing TS Bases states that ``The OPERABILITY of the 
Service Water System ensures that sufficient cooling capacity is 
available for continued operation of safety-related equipment during 
normal and accident conditions. The redundant cooling capacity of 
this system, assuming a single failure, is consistent with the 
assumptions used in the safety analyses.''
    Since this change continues to control the availability of the 
SW pumps by placing the system in an ACTION statement with one loop 
out of service, then the change will continue to comply with the 
existing BASES requirements. Thus it is concluded that the proposed 
revision does not involve a significant reduction in the margin of 
safety.
    In conclusion, based on the information provided, it is 
determined that the proposed revision does not involve a 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 Section Chief: James W. Clifford.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Dockets 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: December 24, 1998.
    Description of amendment request: Revises the setpoints and limits 
of allowable values for loss of power (LOP) instrumentation for 4kV 
emergency busses.
    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 LOP instrumentation provides safety-related electrical 
equipment protection. No new equipment is added to the plant as a 
result of the proposed changes. Separation of the 4kV emergency 
buses from the grid is the only potential transient that previously 
existed based on operation of these relays. Based on the revised 
Voltage Regulation Study, which incorporates the effects of system 
improvements and additional conservatisms, there is no significant 
increase in the probability of this separation. The relay time delay 
settings are such that the relays will detect and respond to an 
actual sustained degradation of voltage, but will not actuate in 
response to normal operational voltage fluctuations. No accident 
initiators will be impacted by the proposed setpoint changes. All 
safety systems will be able to perform their safety functions. 
Accident mitigation is achieved by these relays by ensuring adequate 
voltage is maintained throughout the Class 1E electrical 
distribution system.
    The existing allowable values and the proposed allowable values 
for Functions 2, 3, 4, and 5 have been analyzed and both values are 
acceptable for operation. During implementation of modification 96-
01511 (changing of the relay setpoints), the 4kV buses could be in 
one of the three configurations: (a) Both sources have relays set at 
the existing setpoints, (b) one set of source relays with the 
existing old setpoints and the other set with the proposed revised 
setpoints, or (c) both sources have relays set at the proposed 
revised setpoints. Each of these configurations is acceptable 
because the existing and proposed values satisfy the design limits 
established within the setpoint calculation and the Voltage 
Regulation Study.
    For Function[s] 4 and 5, the present TS has separate entries in 
Table 3.3.8.1-1, for the internal and external time delay. This 
proposed change will combine these internal and external time delays 
for simplicity. The aggregate time delay is the important parameter 
and it is the only time delay that is analyzed. The internal time 
delay minimizes the relay contact wear and reduces the number of 
external time delay relay actuations due to transient voltage dips. 
The internal time delay provides no other output functions. 
Therefore, there will be no impact on the Class 1E power 
distribution system to perform its intended design function.
    Therefore, the proposed changes described above, or operation 
while modification 96-01511 is being implemented, does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed LOP instrumentation setpoint changes will not 
result in any new accidents or operational transients. Separation of 
the 4kV emergency buses from the grid is the only potential 
transient that previously existed based on operation of these 
relays. Based on the revised Voltage Regulation Study, which 
incorporates the effects of system improvements and additional 
conservatisms, there is no significant increase in the probability 
of this separation, and the proposed setpoint changes would not 
create the possibility of a new or different kind of accident from 
any previously evaluated. The relay time delay settings are such 
that the relays will detect and respond to an actual sustained 
degradation of voltage, but will not actuate in response to normal 
operational voltage fluctuations. The proposed setpoint changes for 
these relays and the proposed combining

[[Page 24200]]

of the internal and external time delays will not become initiators 
of different types of accidents or transients. Additionally, since 
the existing and proposed allowable values for the LOP 
instrumentation functions are within the band established by the 
Voltage Regulation Study, both values are acceptable for operation 
during the implementation of modification 96-01511. Therefore, the 
possibility of a new or different kind of accident than previously 
evaluated is not created.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    All LOP instrumentation functions will continue to be carried 
out. The proposed setpoint and allowable value changes have been 
evaluated within the Voltage Regulation Study and the Plant 
Electrical Load Study. The relay setpoints have been established 
using IISCP setpoint methodology. The setpoint determination 
accounts for relay accuracy, potential transformer accuracy, 
measurement and test equipment accuracy, and margin above the design 
limit established within the Voltage Regulation Study. The proposed 
setpoint changes for these relays and the proposed combining of the 
internal and external time delays will not involve a significant 
reduction in a margin of safety. Additionally, since the existing 
and proposed allowable values for the LOP instrumentation functions 
are within the band established by the Voltage Regulation Study, 
both values are acceptable for operation during the implementation 
of modification 96-01511. Therefore, having both values during the 
implementation of modification 96-01511 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: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.
    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: Elinor G. Adensam

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Dockets 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: February 12, 1999.
    Description of amendment request: Administrative changes to correct 
typographic errors in Technical Specifications (TS) introduced in 
previous amendments.
    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 correct typographical errors and are 
administrative only and do not impact the operation of the facility. 
In each case, the action of the intended TS requirements were 
satisfactorily completed when the change was implemented. These 
corrections are administrative only and have no effect on any 
previously evaluated accident scenario. The changes will not alter 
the operation of equipment assumed to be available for the 
mitigation of accidents or transients, nor will they alter the 
operation of equipment important to safety previously evaluated.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes correct typographical errors and are 
administrative only and will not involve any physical changes to the 
plant SSCs [systems, structures, or components]. In each case, the 
action of the intended TS requirements were satisfactorily completed 
when the change was implemented. These corrections are 
administrative only and have no effect on any previously evaluated 
accident scenario. The proposed changes do not allow operation in 
any mode that is not already evaluated. The changes will not alter 
the operation of equipment important to safety previously evaluated.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes correct typographical errors and are 
administrative only and will not affect the manner in which the 
facility is operated, or change equipment or features which affect 
the operational characteristics of the facility. The proposed 
changes have no impact on any safety analysis assumptions or margins 
of safety.
    Therefore, these proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.
    Attorney for Licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101.
    NRC Section Chief: James W. Clifford.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia .

    Date of amendment request: January 21, 1999.
    Description of amendment request: The proposed amendments would 
change Technical Specification Tables 3.3.6.1-1 and 3.3.6.2-1 by 
increasing the Allowable Values for the high radiation trip for the 
exhaust monitors for the reactor building and the refueling. The 
January 21, 1999, amendment request supercedes the July 22, 1998, 
amendment request which was noticed in the Federal Register on August 
26, 1998 (63 FR 45529).
    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). Do the proposed changes involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The Unit 1 and Unit 2 reactor building and refueling floor 
ventilation exhaust radiation monitors perform no function in 
preventing, or decreasing the probability of, a previously evaluated 
accident. The monitors are designed to monitor ventilation exhaust 
for indications of a release of radioactive material resulting from 
a design basis accident and initiate appropriate protective actions. 
Because the proposed changes affect only the ventilation exhaust 
radiation monitors, the probability of an accident previously 
evaluated remains the same.
    The function of the reactor building and the refueling floor 
ventilation exhaust radiation monitors, in combination with other 
accident mitigation systems, is to limit fission product release 
during and following postulated design basis accidents. The proposed 
new Allowable Values for the high radiation trip will continue to 
ensure the offsite doses resulting from a design basis accident do 
not exceed the NRC-approved

[[Page 24201]]

licensing basis. Therefore, the proposed changes do not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes increase the radiation level at which the 
ventilation exhaust monitors actuate; however, the manner in which 
their actuation logic functions and the systems that isolate or 
actuate as a result are unaffected by the proposed changes. 
Furthermore, the ventilation exhaust monitors will continue to 
perform their design function of limiting offsite doses to NRC-
approved licensing limits at the higher Allowable Values. Therefore, 
the proposed changes cannot create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The Bases for Unit 1 and Unit 2 Technical Specifications Tables 
3.3.6.1-1 and 3.3.6.2-1 state that the Allowable Values for the 
reactor building and refueling floor ventilation exhaust radiation 
monitors ``are chosen to ensure radioactive releases do not exceed 
offsite dose limits.'' The proposed Allowable Values ensure the 
radiation monitors actuate at a radiation level sufficient to ensure 
offsite doses are within the NRC-approved licensing basis. The 
proposed Allowable Values comply with the margin of safety defined 
in the Technical Specifications Bases for the ventilation exhaust 
radiation monitors; therefore, the proposed changes do not reduce 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: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC.
    NRC Section Chief: Richard L. Emch, Jr.

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: March 30, 1999.
    Description of amendment request: The licensee has proposed to 
relocate Technical Specification 3/4.3.3.4, ``Meteorological 
Instrumentation,'' and its associated Bases to the Technical 
Requirements Manual (TRM). Because the TRM is incorporated within the 
South Texas Project updated final safety analysis report (UFSAR) for 
the units, changes to the requirements on the meteorological 
instrumentation that would be relocated to the TRM would be controlled 
in accordance with 10 CFR 50.59.
    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 affected system and components [i.e., meteorological 
monitoring instrumentation] are not assumed as initiators of 
analyzed events, and are not assumed to mitigate accident or 
transient events. The requirements and surveillances for [this 
affected system] and components will be relocated from the Technical 
Specifications to the Technical Requirements Manual, which is 
incorporated in the South Texas Project UFSAR and will be maintained 
pursuant to 10 CFR 50.59. In addition, the Meteorological Monitoring 
System components are addressed in existing surveillance procedures 
which are also controlled by 10 CFR 50.59 and subject to the change 
control provisions imposed by plant administrative procedures, which 
endorse applicable regulations and standards. The associated changes 
to the Technical Specification Index are administrative. Therefore, 
this change does 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.
    This change does not involve a physical alteration of the plant 
(no new or different type of equipment will be installed) or make 
changes in the methods governing normal plant operation. This change 
will not impose different requirements, and adequate control of 
information will be maintained. Furthermore, this change will not 
alter assumptions stated in the safety analysis or licensing basis. 
The associated changes to the Technical Specification Index are 
administrative. Therefore, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    This change will not reduce a margin of safety because the 
change has no impact on any safety analysis assumptions. In 
addition, the relocated requirements and surveillances for the 
affected structures, systems, and components remain the same as the 
existing Technical Specifications. Because any future changes to 
these requirements or the surveillance procedures will be evaluated 
per the requirements of 10 CFR 50.59; there is no [significant] 
reduction in a margin of safety. The associated changes to the 
Technical Specification Index are administrative and have no 
potential effect on the margin of safety. Therefore, this change 
does not involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 
77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

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: March 2, 1999 (TS 98-05).
    Brief description of amendments: The proposed amendments would 
change the SQN Operating Licenses DPR-77 (Unit 1) and DPR-79 (Unit 2) 
by eliminating a requirement to have an Independent Safety Engineering 
Group (ISEG), conditions imposed by NUREG-0737. Because of evolution 
through numerous reorganizations and reassignments, these license 
conditions are no longer necessary and the Tennessee Valley Authority 
(TVA, the licensee) proposes deleting them.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The possibility of occurrence or the consequences for an 
accident or malfunction of equipment is not increased. The ISEG 
function is one of ``oversight'' only.
    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 Final Safety Analysis Report 
is not created by the proposed elimination of the ISEG; nor is the 
possibility for an accident or malfunction of a different type. The 
ISEG function is one of ``oversight'' only.

[[Page 24202]]

    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed amendment will not involve a significant reduction 
in the margin of safety. The ISEG function is one of ``oversight'' 
only.

    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 Section Chief: Sheri R. Peterson.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: February 12, 1999 (TXX-99022).
    Brief description of amendments: The proposed changes would modify 
the steam generator tube inspection requirements and acceptance 
criteria to implement the 1.0-volt repair criteria for steam generator 
tubes affected by outer diameter stress corrosion cracking (ODSCC) 
according to Nuclear Regulatory Commission (NRC) Generic Letter 95-05 
(``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes 
Affected by Outside Diameter Stress Corrosion Cracking'') at Comanche 
Peak Unit 1. Also proposed is the use of a voltage-dependent 
probability of detection; the methodology was originally submitted to 
the NRC by the Nuclear Energy Institute in 1996.
    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 Comanche Peak Unit 1 in accordance with the 
proposed license amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Tube burst criteria are inherently satisfied during normal 
operating conditions due to the proximity of the tube support plate 
[TSP]. Test data indicates that tube burst cannot occur within the 
TSP, even for tubes which have 100% through-wall electric discharge 
machining notches, 0.75 inch long, provided that the TSP is adjacent 
to the notched area. Since tube to tube support plate proximity 
precludes tube burst during normal operating conditions, use of the 
criteria must retain tube integrity characteristics which maintain a 
margin of safety of 1.43 times the bounding faulted condition (Steam 
Line Break) pressure differential. As previously stated, the RG 
[Regulatory Guide] 1.121 criterion requiring maintenance of a safety 
factor of 1.43 times the Steam Line Break pressure differential on 
tube burst is satisfied by \3/4\'' diameter tubing with bobbin coil 
indications with signal amplitudes less than 4.7 volts, regardless 
of the indicated depth measurement. At the FDB [flow distribution 
baffle], a safety factor of 3 against the normal operating condition 
at power is applied. Here a voltage of 3.34 volts satisfies the 
burst capability recommendation.
    The upper voltage repair limit (VURL) will be 
determined prior to each outage using the most recently approved NRC 
database to determine the tube structural limit (VSL). 
The structural limit is reduced by allowances for nondestructive 
examination (NDE) uncertainty (VNDE) and growth 
(VGr) to establish VURL. As an example, the 
NDE uncertainty component of 20% and a voltage growth allowance of 
30% per full power year can be utilized to establish a 
VURL of 3.13 volts for TSP indications, 2.22 volts for 
the FDB indications. The 20% NDE uncertainty represents a 
squareroot-sum-of-the-squares (SRSS) combination of probe wear 
uncertainty and analyst variability.
    The flaw growth allowance should be an average growth rate or 
30% per effective full power year, whichever is larger. The 30% 
growth allowance used to determine VURL is conservative 
for the current conditions at Comanche Peak Unit 1. The average 
growth of the bobbin indication voltages observed at the last 
inspection is determined to be 0.14 volts, or 24.6% voltage growth. 
This value is a conservative representation of the growth trends at 
Comanche Peak Unit 1 as not all steam generators were inspected at 
end of cycle 3 and end of cycle 4, and the largest reported voltage 
growths represent more than one cycle of actual plant operation. The 
most current NRC approved database, contained in EPRI [Electric 
Power Research Institute] NP-7480-L, Addendum 1, was used to 
establish the VURL values for the FDB and TSP 
intersections. Once approved by the NRC, the industry protocol for 
updating the database will be followed by TU Electric, ensuring that 
the most current database is utilized for all future applications of 
the criteria.
    Also, assuming the criteria was applied at the last inspection 
at Comanche Peak Unit 1, using conservative growth projections as 
described in Reference 2 [of the February 12, 1999, application], 
the conditional burst probability at end of cycle 6 is determined to 
be 1.7  x  10-4, which is well within the GL 95-05 
reporting limit of 1 x 10-2.
    Relative to the expected leakage during accident condition 
loadings, it has been previously established that a postulated main 
Steam Line Break outside of containment but upstream of the MSIV 
[main steam isolation valve] represents the most limiting 
radiological condition relative to the plugging criteria. In support 
of implementation of the revised plugging limit, it will be 
determined whether the distribution of cracking indications at the 
tube support plate intersections during future cycles are projected 
to be such that primary to secondary leakage would result in site 
boundary doses within 10CFR100 guidelines and control room doses 
within the GDC [General Design Criterion]-19 limit. A separate 
calculation has determined this allowable Steam Line Break leakage 
limit to be 27.79 gpm in the faulted loop assuming a RCS [reactor 
coolant system] dose equivalent I-131 concentration of 1.0 microCi/
gm. The establishment of the 27.79 gpm leak rate value is controlled 
by the 0 to 2 hour offsite dose at the site boundary for the 
accident initiated iodine spike case, not the control room dose. For 
this case, the site boundary thyroid dose approaches, but is bounded 
by, the 30 Rem limit recommended in NUREG-0800 [``Standard Review 
Plan''].
    The methods for calculating the radiological dose consequences 
are also revised for this application. Rather than basing the 
calculated thyroid dose consequences on conversion factors from TID-
14844, [``Calculation of Distance Factors for Power and Test Reactor 
Sites''] factors obtained from ICRP-30 [International Commission on 
Radiation Protection Publication 30] are used. The use of ICRP-30 
dose conversion factors in this application has been previously 
accepted by the NRC. Although the use of ICRP-30, relative to the 
TID-14844, results in lower calculated thyroid doses for this 
application, the NRC has previously determined that the ICRP-30 
factors retain adequate conservatism.
    In summary, due to the methodology used to determine the maximum 
allowable, accident-initiated leak rate (prescribed in Section 2.b.4 
of Generic Letter 95-05), the calculated radiological consequences 
at the EAB [exclusion area boundary] and LPZ [low population zone] 
are larger than previously reported for the postulated steamline 
break event. However, the calculated radiological consequences 
remain in compliance with NUREG-0800 and GDC-19. Therefore, it is 
concluded that the proposed changes do not result in a significant 
increase in the radiological consequences of an accident previously 
analyzed.
    The removal from the FSAR [final safety analysis report] of the 
steamline break radiological dose consequences calculation typically 
identified as a ``5% failed fuel'' scenario does not affect the 
probability or consequences of any accident previously considered. 
For CPSES [Comanche Peak Steam Electric Station], no accident-
induced fuel failures are predicted; therefore, consistent with 
NUREG-0800, this scenario is not required to be analyzed or 
presented in the FSAR.
    In summary, because the implementation of the 1.0 volt voltage-
based plugging criteria at Comanche Peak Unit 1 does not adversely 
affect steam generator tube integrity and implementation will be 
shown to result in acceptable radiological dose consequences, the 
proposed Technical Specification change does not result in any 
increase in the probability or consequences of an accident 
previously evaluated within the Comanche Peak FSAR.
    (2) The proposed license amendment does not create the 
possibility of a new or different

[[Page 24203]]

kind of accident from any accident previously evaluated.
    Implementation of the proposed steam generator tube 1.0 volt 
plugging limit does not introduce any significant changes to the 
plant design basis. Neither a single or multiple tube rupture event 
would be expected in a steam generator in which the plugging limit 
has been applied (during all plant conditions).
    The bobbin probe voltage-based tube plugging criteria of 1.0 
volt is supplemented by: enhanced eddy current inspection guidelines 
to provide consistency in voltage normalization, a 100% eddy current 
inspection sample size at the tube support plate elevations, and RPC 
[rotating pancake coil] inspection requirements for the larger 
indications left in service to characterize the principal 
degradation as ODSCC. TU Electric will implement a maximum normal 
operating condition primary to secondary leakage rate limit of 150 
gpd (0.1 gpm--at room temperature) per steam generator to help 
preclude the potential for excessive leakage during all plant 
conditions. The 150 gpd leakage limit is more restrictive than the 
standard operating leakage limit (of 500 gpd) and is intended to 
provide additional margin to accommodate a stress corrosion crack 
which might grow at a greater than expected rate or unexpectedly 
extend outside the thickness of the tube support plate. Leakage 
trending capability consistent with EPRI Report TR-04788, ``PWR 
Primary-to-Secondary Leak Guidelines'', has been implemented at 
Comanche Peak Unit 1.
    As steam generator tube integrity upon implementation of the 1.0 
volt plugging limit continues to be maintained through in-service 
inspection and primary to secondary leakage monitoring, the 
possibility of a new or different kind of accident from any accident 
previously evaluated is not created.
    (3) The proposed license amendment does not involve a 
significant reduction in margin of safety.
    The use of the voltage-based bobbin probe tube support plate 
elevation plugging criteria at Comanche Peak Unit 1 maintains steam 
generator tube integrity commensurate with the criteria of 
Regulatory Guide 1.121. Regulatory Guide 1.121 describes a method 
acceptable to the NRC staff for meeting GDCs 14, 15, 31, and 32 by 
reducing the probability or the consequences of steam generator tube 
rupture. This is accomplished by determining the limiting conditions 
of degradation of steam generator tubing, as established by 
inservice inspection, for which tubes with unacceptable cracking 
should be removed from service. Upon implementation of the proposed 
criteria, even under the worst case conditions, the occurrence of 
ODSCC at the tube support plate elevations is not expected to lead 
to a steam generator tube rupture event during normal or faulted 
plant conditions. The end of cycle distribution of crack indications 
at the tube support plate elevations is confirmed to result in 
acceptable primary to secondary leakage during all plant conditions 
and that radiological consequences are not adversely 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: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019 Attorney for licensee: George L. Edgar, 
Esq., Morgan, Lewis and Bockius, 1800 M Street, NW., Washington, DC 
20036.
    NRC Section Chief: Robert A. Gramm
    Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.

    Date of amendment request: February 16, 1999.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) Sections 3.6, 3.9, and 3.16 
and the associated Bases for those sections for Units 1 and 2. The 
proposed changes would consolidate the auxiliary feedwater (AFW) cross-
connect requirements by relocating the electrical power requirements 
from Section 3.16 to Section 3.6. The proposal also would clarify the 
TS with regard to permitting simultaneous entry into certain conditions 
of operation on Units 1 and 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:

    Criterion 1--Operation of Surry Units 1 and 2 in accordance with 
the proposed TS change does not involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated.
    The proposed TS change is administrative in nature, and station 
operations are not being affected. The accidents considered relative 
to this proposed TS change are Rupture of Main Steam Pipe, Loss of 
All AC Power, and Loss of Feedwater. The probability of occurrence 
of these accidents has been previously evaluated to support Surry TS 
Amendment 143/140. The NRC reviewed the PSA [probabilistic safety 
analysis] basis during issuance of TS Amendment 143/140 and found it 
acceptable. The probability of occurrence of these accidents has 
been recently reviewed relative to this proposed TS change. It has 
been concluded that the proposed TS change is consistent with the 
existing analyses and evaluations and, therefore, will not increase 
the probability of occurrence of the identified accidents.
    The consequences of the accidents identified above were also 
previously evaluated to support Surry TS Amendment 143/140. The PSA 
considerations included the AFW cross-connect capability, diesel 
generator dependencies, various LCO [limiting condition for 
operation] time periods, and a HELB [high energy line break] in the 
vicinity of the AFW Pumps. The previous evaluation was recently 
reviewed relative to this proposed TS change. This review determined 
that the proposed TS change is consistent with the design and 
licensing bases supporting the existing Technical Specifications. 
The proposed TS change is also consistent with the existing analyses 
and evaluations, the consequences of which bound any potential 
consequences of the proposed TS change. Therefore, the proposed TS 
change will not increase the consequences of the identified 
accidents.
    Criterion 2--The proposed TS change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The possibility for a new or different type of accident than any 
previously evaluated is not created since the considerations in the 
PSA and evaluations performed to support TS Amendment 143/140 are 
not changed by the proposed administrative TS change. The proposed 
TS change is consistent with the design and licensing bases 
supporting the existing Technical Specifications. Furthermore, 
station operations and plant equipment are not being affected and, 
therefore, the proposed TS change does not create any new failure 
modes or accident precursors.
    Criterion 3--The proposed TS change does not involve a 
significant reduction in a margin of safety.
    The proposed administrative change to Surry Technical 
Specifications clarifies the requirements (limiting conditions for 
operation (LCO) and action statements) relating to the Auxiliary 
Feedwater (AFW) cross-connect by relocating the emergency power 
source requirements of TSs 3.16.A.8 and 3.16.B.4 to TS 3.6. The 
proposed TS change does not alter the current TS requirements or 
bases, as well as maintains the Surry licensing and design basis. 
The proposed change does not affect either station operations or 
plant equipment, hence the availability of equipment for the 
mitigation of accidents is not decreased. Furthermore, the 
assumptions governing the accident analyses remain unchanged, and 
the consequences of the existing analyses and evaluations remain 
bounding. This is an administrative change and as such does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.

[[Page 24204]]

    Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams, 
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
23219.
    NRC Section Chief: Richard L. Emch, Jr.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: February 16, 1999.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) Section 4.2 for Units 1 and 2 
to relax the surveillance requirements for reactor coolant pump (RCP) 
flywheels. The flywheels provide extended reactor coolant flow 
coastdown capability if electric power for the RCPs is lost. Currently, 
the flywheels are subjected to an inspection program that meets the 
requirements of NRC Regulatory Guide 1.14, Revision 1, dated August 
1975. The inspections include an ultrasonic examination (UT) of areas 
of high stress concentration at the bore and keyway every three years, 
and complete UT every 10 years. The proposed change would require only 
a 10-year UT, based upon an analysis presented in a Westinghouse 
topical report (WCAP-14535A) which has been reviewed and accepted by 
NRC staff.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    a. The reduction of the inspection requirements for the reactor 
coolant pump flywheels, as generically approved by the NRC and 
technically supported by WCAP-14535A, does not significantly 
increase the probability of an accident previously evaluated in the 
safety analysis report. The results of WCAP-14535A have been 
reviewed and evaluated with the technical basis accepted for 
referencing in license applications by the NRC in their letter 
entitled ``Acceptance for referencing of Topical Report WCAP-14535, 
Topical Report on Reactor Coolant Pump Flywheel Inspection 
Elimination,'' dated September 12, 1996.
    The proposed Technical Specification change reduces the 
surveillance requirements (inspection) on the RCP flywheel. There is 
no change in the method of plant operation or system design. The 
WCAP-14535A report establishes that the proposed change has a 
negligible affect on the probability that the flywheel will fail 
given that the flywheels received preservice and inservice 
examinations as required previously. Therefore, the proposed change 
does not increase the probability of occurrence or consequences of 
any previously analyzed accident.
    b. The proposed change to reduce the inspection requirements for 
the RCP flywheels as generically approved by the NRC and supported 
by WCAP-14535A does not create the possibility of a new or different 
kind of accident from any accident previously evaluated in the 
safety analysis report.
    The proposed surveillance requirements (inspection) only reduce 
the inspection requirements/frequency for the reactor coolant pump 
flywheels, and there is no change in the method of plant operation 
or system design.
    c. The proposed change reducing the inspection of the RCP 
flywheels as generically approved by the NRC and supported by WCAP-
14535A, does not impact the accident analysis assumptions or the 
basis of any Technical Specification. As previously stated, the 
analysis performed in the WCAP-14535A report established that the 
affect on flywheel failure probability was negligible given that the 
initial preservice and inservice inspections under the current 
requirements were performed. Therefore, the proposed change in 
surveillance (inspection) frequency does not involve a significant 
reduction in the margin of safety.
    The analysis provided herein demonstrates that the proposed 
amendment to the Surry Technical Specifications does not involve a 
significant increase in the probability or consequences of a 
previously evaluated accident, does not create the possibility of a 
new or different kind of accident, and does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams, 
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
23219.
    NRC Section Chief: Richard L. Emch, Jr.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of amendment request: April 12, 1999 (TSCR 212).
    Description of amendment request: The purpose of the proposed 
amendments is to update references in the Technical Specifications. The 
update is necessary to reflect relocation of referenced information in 
the Final Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed amendment corrects references within the Technical 
Specification requirements such that they refer to the correct 
information in the updated Final Safety Analysis Report (FSAR). The 
references changed due to relocation of the information within the 
FSAR. The Technical Specification requirements and intent are not 
changed. Therefore, these changes are administrative only and do not 
change the design or operation of the Point Beach Nuclear Plant 
[PBNP]. Operation of PBNP in accordance with the proposed amendments 
cannot increase the probability or consequences of an accident 
previously evaluated.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create the possibility of a new or 
different kind of accident previously evaluated.
    The proposed changes are administrative only and therefore do 
not materially change any requirements for the design or operation 
of PBNP. Therefore, operation in accordance with the proposed 
changes cannot create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not create a significant reduction in a 
margin of safety.
    The proposed changes are administrative only; correcting 
references within the Technical Specification requirements. No 
requirement on the operation or design of the facility is being 
changed. Therefore, there is no 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: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: George F. Dick, Jr., Acting.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the

[[Page 24205]]

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

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of application for amendments: November 19, 1998.
    Brief description of amendments: The amendments revised Technical 
Specification 3.7.6 ``Service Water (SRW) System'' to allow operation 
of Calvert Cliffs Unit Nos. 1 and 2 with one SRW plate and frame heat 
exchanger in a subsystem secured and removing one containment air 
cooler from service to enable the affected SRW subsystem to remain 
operable.
    Date of issuance: April 14, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 230 and 206.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 16, 1998 (63 
FR 69333). The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated April 14, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of amendment request: November 1, 1996, as supplemented May 
22, 1998, September 14, 1998, January 4, 1999, and March 19, 1999.
    Brief description of amendment: The amendment modified the 
Technical Specifications for the Brunswick Steam Electric Plant, Units 
1 and 2, to extend the Allowed Outage Time for 4.16kV AC balance of 
plant buses and the AC electrical power distribution system load group 
buses.
    Date of issuance: April 15, 1999.
    Effective date: April 15, 1999.
    Amendment Nos.: 205 and 235.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendment 
revises the Technical Specifications.
    Date of initial notice in Federal Register: February 11, 1998 (63 
FR 6977). The supplemental submittals of May 22, 1998, September 14, 
1998, January 4, 1999, and March 19, 1999, contained clarifying 
information only, and did not change the initial no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 15, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: October 14, 1998.
    Brief description of amendment: The amendment modifies the 
acceptance criterion for Surveillance Requirement 3.4.14.2 from the 
setpoint value of 465 psig to the analytical limit for the residual 
heat removal system of 474 psig reactor coolant system pressure.
    Date of issuance: April 20, 1999.
    Effective date: April 20, 1999.
    Amendment No. 182.
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 4, 1998 (63 FR 
59587).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 20, 1999.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of application for amendment: September 3, 1997, as 
supplemented March 13, 1998, and March 18, 1999.
    Brief description of amendment: The amendment revises the technical 
specifications to delete snubber operability requirements, action 
requirements for inoperable snubbers, and snubber testing requirements. 
The snubber testing requirements have been relocated to the Palisades 
Operating Requirements Manual.
    Date of issuance: April 13, 1999.
    Effective date: April 13, 1999, and shall be implemented within 60 
days.
    Amendment No.: 185.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 8, 1998 (63 FR 
17222). The March 18, 1999, submittal requested a 60-day allowance for 
implementation of the amendment. This change was within the scope of 
the original Federal Register notice and did not change the staff's 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 13, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423-3698.

Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power 
Plant, Unit 1, Monroe County, Michigan

    Date of amendment request: July 17, 1998 (Reference NRC-98-0044).
    Brief description of amendment: This amendment revises the Enrico 
Fermi

[[Page 24206]]

Atomic Power Plant, Unit 1, License to allow possession of a nominal 
amount of special nuclear material.
    Date of issuance: April 15, 1999.
    Effective date: On the date of issuance of this amendment and must 
be fully implemented no later than 60-calendar days from the date of 
issuance.
    Amendment No.: 16.
    Facility Operating License No. DPR-9: Amendment revised the License 
by adding new Part 2.B.4 to the License.
    Date of initial notice in Federal Register: October 21, 1998 (63 FR 
56240). The NRC's related evaluation of the amendment is contained in a 
Safety Evaluation dated April 15, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.

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

    Date of application for amendments: March 15, 1999, and 
supplemented by letter dated March 17, 1999.
    Brief description of amendments: The amendments delete from the 
joint Technical Specifications Section 3.3.7, ``Control Room Area 
Ventilation System (CRAVS) Actuation Instrumentation,'' and Section 
3.3.8, ``Auxiliary Building Filtered Ventilation Exhaust System 
(ABFVES) Actuation Instrumentation.'' These surveillance requirements 
are not applicable to Catawba because the sections do not reflect the 
design of the Catawba units.
    Date of issuance: April 8, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--177; Unit 2--169.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in FEDERAL REGISTER: March 24, 1999 (64 FR 
14274). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

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

    Date of application for amendments: February 18, 1999.
    Brief description of amendments: The amendments revise the 
Technical Specifications Surveillance Requirement (SR) 3.6.16.1 
regarding surveillance of reactor building access openings, SR 3.6.16.3 
regarding surveillance of reactor building structural integrity, and 
Administrative Controls 5.5.2 regarding the Containment Leakage Rate 
Testing Program. The revised requirements would provide scheduling 
flexibility without decreasing quality and safety margin.
    Date of issuance: April 9, 1999.
    Effective date: As of the date of issuance, to be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 178--Unit 1; 170--Unit 2.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 10, 1999 (64 FR 
11961). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 9, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley Power 
Station, Unit No. 1, Shippingport, Pennsylvania

    Date of application for amendment: November 11, 1998, as 
supplemented February 26, 1999.
    Brief description of amendment: The amendment modified License 
Condition 2.C(9) to allow, on a one-time only, extension of the steam 
generator inspection interval in Technical Specification Surveillance 
4.4.5.3.b. This will allow the steam generator inspection interval to 
coincide with the thirteenth refueling outage or the end of 500 
effective full power days, whichever occurs sooner.
    Date of issuance: April 16, 1999.
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment No: 221.
    Facility Operating License No. DPR-66. Amendment revised the 
License.
    Date of initial notice in Federal Register: December 2, 1998 (63 FR 
66593). The February 26, 1999, letter provided additional information 
but did not change the initial proposed no significant hazards 
consideration determination or expand the amendment request beyond the 
scope of the initial notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 16, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B.F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Units 1 and 2, Pope County, Arkansas

    Date of amendment request: June 28, 1996, as supplemented by 
letters dated February 23 and March 15, 1999.
    Brief description of amendments: The amendments revise the 
Technical Specifications to permit the containment equipment hatch to 
be open during handling of irradiated fuel in containment and core 
alterations provided that the capability for closure is maintained.
    Date of issuance: April 16, 1999.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days of issuance.
    Amendment Nos.: Unit 1--Amendment No. 195; Unit 2--Amendment No. 
203.
    Facility Operating License Nos. DPR-51 and NPF-6: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 14, 1996 (61 FR 
42280). The February 23 and March 15, 1999, letters provided clarifying 
information that did not change the scope of the original application 
and the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 16, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of application for amendment: April 30, 1998.
    Brief description of amendment: The amendment revises the single 
largest post-accident load capable of being supplied by the diesel 
generators and relocates this value to the Bases for Technical 
Specification (TS)

[[Page 24207]]

Surveillance 4.8.1.1.2.c.3. TS Surveillance 4.8.1.1.2.c.3 has been 
revised to refer to ``the single largest post-accident load'' rather 
than a specific numerical value for diesel generator load reject 
testing. This change is consistent with the guidance provided in NUREG-
1432 , ``Improved Standard Technical Specifications for Combustion 
Engineering Plants.''
    Date of issuance: April 21, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 204.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 21, 1998 (63 FR 
56241). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 21, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: November 13, 1997.
    Brief description of amendment: The amendment changes the Appendix 
A Technical Specifications (TSs) by revising TS 6.8.4.a, Primary 
Coolant Sources Outside Containment, to add portions of the containment 
vacuum relief and primary sampling systems to the list of systems 
included in the Primary Coolant Sources Outside Containment Program.
    Date of issuance: April 21, 1999.
    Effective date: The license amendment is effective as of its date 
of issuance, and shall be implemented within 60 days.
    Amendment No.: 150.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 25, 1998 (63 
FR 9601). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 21, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit 3, Citrus County, Florida

    Date of application for amendment: October 1, 1997, as supplemented 
April 23 and November 17, 1998 and February 19, 1999.
    Brief description of amendment: The changes specify criteria for 
evaluating the growth of pit-like intergranular attack steam generator 
tube degradation identified in tubes in the ``B'' once-through steam 
generator (OTSG). Florida Power Corporation also requested to amend the 
Improved Technical Specifications to clarify the date by which the OTSG 
inservice inspection results are required to be submitted to the NRC.
    Date of issuance: April 8, 1999.
    Effective date: April 8, 1999.
    Amendment No.: 172.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 22, 1997 (62 FR 
54873). The supplemental letters dated April 23 and November 17, 1998, 
and February 19, 1999 did not change the original no significant 
hazards determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit 3, Citrus County, Florida

    Date of application for amendment: September 9, 1997, as 
supplemented November 7 and 25, 1997, and January 20 and October 30, 
1998.
    Brief description of amendment: The amendment proposed to revise 
the Final Safety Analysis Report (FSAR) analysis of the Makeup System 
letdown line failure accident. The revised analysis models the event as 
being terminated by manual operator action to isolate the line whereas 
the original analysis models an automatic isolation of the break.
    Date of issuance: April 13, 1999.
    Effective date: April 13, 1999.
    Amendment No.: 173.
    Facility Operating License No. DPR-72: Amendment approves changes 
to the Final Safety Analysis Report.
    Date of initial notice in Federal Register: September 24, 1997 (62 
FR 50005). The supplemental letters dated November 7 and 25, 1997, 
January 20, 1998, and October 30, 1998, did not change the original 
proposed no significant hazards consideration determination, or expand 
the scope of the amendment request as originally noticed.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 13, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit 3, Citrus County, Florida

    Date of application for amendment: October 30, 1998, as 
supplemented April 7, 1999.
    Brief description of amendment: Changes the Crystal River Unit 3 
Technical Specifications to delete a note regarding the number of 
required channels for the Degrees of Subcooling function, and to 
subdivide the Core Exit Temperature (Backup) function into two new 
functions in Table 3.3.17-1, Post-Accident Monitoring Instrumentation.
    Date of issuance: April 20, 1999.
    Effective date: April 20, 1999.
    Amendment No.: 174.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 13, 1999 (64 FR 
2246). The April 7, 1999, supplement did not affect the original no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 20, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No. 2, St. Lucie County, Florida

    Date of application for amendment: May 27, 1998, as supplemented 
October 9, 1998.
    Brief description of amendment: Deletes the requirement for 
operability of the safety injection tanks in Mode 4 of reactor 
operation.
    Date of Issuance: April 8, 1999.
    Effective Date: Amendment is effective within 30 days of receipt.
    Amendment No.: 100.

[[Page 24208]]

    Facility Operating License No. NPF-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 29, 1998 (63 FR 
40556). The October 9, 1998 supplemental letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear 
Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of application for amendment: November 25, 1998, as 
supplemented February 12, 1999.
    Brief description of amendment: The amendment approves the proposed 
surveillance Technical Specifications related to the once through steam 
generator inservice inspections to be completed during the 13R 
refueling outage in fall 1999. Related TS Bases changes are also 
included.
    Date of issuance: April 13, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 209.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 16, 1998 (63 
FR 69342).
    The February 12, 1999, submittal modified the request, but did not 
affect the initial no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 13, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (Regional Depository) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear 
Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of application for amendment: October 15, 1998, as 
supplemented February 3, and February 12, 1999.
    Brief description of amendment: The amendment authorizes a revision 
to the TMI-1 updated final safety analysis report (UFSAR) for use of 
revised atmospheric dispersion factors (X/Q) (obtained by utilizing 
recent meteorological data) in determining Chapter 14 postulated 
accident analysis radiological dose consequences at Technical 
Specification Section 5.1.1 defined exclusion area boundary (EAB) and 
low population zone (LPZ).
    Date of issuance: April 15, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 210.
    Facility Operating License No. DPR-50. Amendment authorizes changes 
to the UFSAR.
    Date of initial notice in Federal Register: November 18, 1999 (63 
FR 64117).
    The February 3, and February 12, 1999, letters were within the 
scope of the original application and did not change the staff's no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 15, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (Regional Depository) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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

    Date of application for amendment: December 28, 1998, as 
supplemented March 1 and 29, 1999.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 2.2.1, ``Limiting Safety System Settings-Reactor 
Trip Setpoints,'' to reflect revised loss of normal feedwater flow 
analyses.
    Date of issuance: April 8, 1999.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 232.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 10, 1999 (64 
FR 6701). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, Attn: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

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

    Date of application for amendment: January 18, 1999.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.6.1.2, ``Containment Systems--Containment 
Leakage,'' and also revises the related TS bases and Final Safety 
Analysis Report sections. The revisions relate to changes in the 
secondary containment bypass leakage.
    Date of issuance: April 14, 1999.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 234.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 10, 1999 (64 
FR 6703). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 14, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, Attn: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone 
Nuclear Power Station, Unit No. 3, New London County, Connecticut

    Date of application for amendment: February 10, 1999.
    Brief description of amendment: The amendment incorporates 
alternative inspection requirements into Technical Specification 
Surveillance Requirement 3/4.4.10, ``Structural Integrity,'' for the 
reactor coolant pump flywheel.
    Date of issuance: April 16, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 169.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 10, 1999 (64 FR 
11964).

[[Page 24209]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 16, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, Attn: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of application for amendments: February 5, 1999, as 
supplemented March 1, 1999.
    Brief description of amendments: The amendments revise certain 
requirements for repair of defective steam generator tubs specified in 
Technical Specification 4.12, ``Steam Generator Tube Surveillance,'' 
based on the latest revision to a previously approved methodology.
    Date of issuance: April 15, 1999.
    Effective date: April 15, 1999, with full implementation within 30 
days.
    Amendment Nos.: 144 Unit 1--135 Unit 2.
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 10, 1999 (64 FR 
11964). The March 1, 1999, supplement provided corrected Technical 
Specification pages. This information was within the scope of the 
original Federal Register notice and did not change the staff's initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 15, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No.1, Washington County, Nebraska

    Date of amendment request: March 18, 1998.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 5.2.f and TS 5.11.2 to change the title of ``Shift 
Supervisor'' to ``Shift Manager.''
    Date of issuance: April 15, 1999.
    Effective date: April 15, 1999.
    Amendment No.: 190.
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 8, 1998 (63 FR 
17227).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 15, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.

Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: May 16, 1996.
    Brief description of amendment: The amendment revises requirements 
for Plant Operating Review Committee review of fire protection program 
and procedure changes.
    Date of issuance: April 12, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 252.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34895).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 12, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: December 16, 1998, as 
supplemented March 22, 1999.
    Brief description of amendment: This amendment revised Technical 
Specification (TS) Surveillance Requirements 4.8.1.1.2 and 4.8.1.1.3, 
Table 4.8.1.1.2-1, and the associated Bases. These changes removed the 
emergency diesel generator accelerated testing and special reporting 
requirements from the TSs in accordance with the guidance provided in 
Generic Letter 94-01.
    Date of issuance: April 14, 1999.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 119.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 13, 1999 (64 FR 
2251).
    The supplemental letters provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 14, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.

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

    Date of application for amendment: June 12, 1998, as supplemented 
July 23, 1998 and September 8, 1998.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Limiting Condition for Operation Sections 3.7.1.1, 
3.7.1.2, and 3.7.1.3. Specifically, the changes revise the Ultimate 
Heat Sink limits for river water temperature, in order to increase 
operational flexibility. In addition, the Station Service Water System 
(SSWS) and Safety Auxiliaries Cooling System (SACS) TS Action 
Statements have been revised to provide additional restrictions on 
continued plant operation. These revisions provide more explicit TS 
direction for plant operation under limiting SSWS/SACS configurations.
    Date of issuance: April 19, 1999.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 120.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 1, 1999 (63 FR 
35995) The July 23, 1998, and September 8, 1998, supplements provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination or expand the scope of 
the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 19, 1999.
    No significant hazards consideration comments received: No

[[Page 24210]]

    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of application for amendment: September 18, 1998, as 
supplemented by letter dated February 5, 1999.
    Brief description of amendment: The amendment revises Virgil C. 
Summer Nuclear Station Technical Specifications to permit use of the 
BEACON system. BEACON is a core power distribution monitoring and 
support system based on a three-dimensional nodal code.
    Date of issuance: April 9, 1999.
    Effective date: April 9, 1999.
    Amendment No.: 142.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 64121).
    The February 5, 1999, submittal contained clarifying information 
only, and did not change the initial no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated April 9, 1999.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San 
Diego County, California

    Date of application for amendments: January 24, 1997.
    Brief description of amendments: The amendments revised 
Surveillance Requirement (SR) 3.8.1.9 to Technical Specification 3.8.1, 
``AC Sources--Operating,'' to more accurately reflect test conditions 
and plant design requirements.
    Date of issuance: April 9, 1999.
    Effective date: April 9, 1999, to be implemented within 30 days 
from the date of issuance.
    Amendment Nos.: Unit 2-151; Unit 3-143.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 11, 1998 (63 
FR 6997).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 9, 1999.
    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: January 20, 1999.
    Brief description of amendments: The amendments revise the 
descriptive details of Technical Specification 4.7.1.2.1.a, regarding 
performance testing of the Auxiliary Feedwater (AFW) pumps, to more 
closely adhere to NUREG-1431, ``Improved Standard Technical 
Specifications for Westinghouse Plants.'' This involves relocating the 
surveillance-required numerical values for the AFW pump performance 
test discharge pressure and flow rate to the South Texas Project 
Updated Final Safety Analysis Report.
    Date of issuance: April 16, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--Amendment No. 105; Unit 2--Amendment No. 
92.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1999 (64 
FR 9201).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 16, 1999.
    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: January 26, 1999.
    Brief description of amendments: The amendments revise part of the 
inservice inspection requirements for the reactor coolant pump flywheel 
from an in-place ultrasonic volumetric examination of the areas of 
higher stress concentration at the bore and keyway at approximately 3-
year intervals and a surface examination of all exposed surfaces and 
complete ultrasonic volumetric examination at approximately 10-year 
intervals to ultrasonic examination over the volume from the inner bore 
of the flywheel to the circle of one-half the outer radius once every 
10 years.
    Date of issuance: April 16, 1999.
    Effective date: April 16, 1999, to be implemented within 30 days of 
issuance.
    Amendment Nos.: Unit 1--Amendment No. 106; Unit 2--Amendment No. 
93.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 10, 1999 (64 FR 
11968).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 16, 1999.
    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, Texas 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: September 30, 1998.
    Brief description of amendments: Revises Units 1 and 2 Technical 
Specification (TS) Section 3/4.4.5, ``Steam Generator'' Surveillance 
Requirements. The future installation of the new Delta 94 steam 
generators at the South Texas Project, Units 1 and 2 necessitates 
changes to the steam generator tube sample selection and inspection 
requirements; inservice inspection frequencies; acceptance criteria; 
and inspection reporting requirements.
    Date of issuance: April 19, 1999.
    Effective date: April 19, 1999, to be implemented following the 
replacement of Unit 1 Model E steam generators with Model delta94 steam 
generators and prior to Unit 1 operation with the delta94 steam 
generators installed.
    Amendment Nos.: Unit 1--Amendment No. 107; Unit 2--Amendment No. 
94.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 4, 1998 (63 FR 
59595).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 19, 1999.
    No significant hazards consideration comments received: No.

[[Page 24211]]

    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: August 6, 1997, as supplemented by 
letters dated September 4 and 18, 1997, December 9, 1997, and February 
4, 1999.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) Table 2.2-1 and TS 3/4.2.5 to allow the reactor 
coolant system total flow rate to be determined using cold leg elbow 
tap differential pressure measurements.
    Date of issuance: April 19, 1999.
    Effective date: As of its date of issuance to be implemented within 
7 days of issuance.
    Amendment Nos.: Unit 1--Amendment No. 108; Unit 2--Amendment No. 
95.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 14, 1997 (62 FR 
43556).
    The September 4 and 18, 1997, December 9, 1997, and February 4, 
1999, letters provided clarifying information that did not change the 
original application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 19, 1999.
    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, Texas 77488.

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant, 
Unit 1, Hamilton County, Tennessee

    Date of application for amendments: August 27, 1998, supplemented 
by letter dated March 19, 1999 (TS 98-04).
    Brief description of amendments: The amendments change the 
Technical Specifications (TS) for Sequoyah Nuclear Plant, Unit 2 
reactor by adding a sentence at the end of TS Section 5.3 authorizing 
installation of a limited number of lead test assemblies containing 
downblended uranium in accordance with Topical Report BAW-2328.
    Date of issuance: April 12, 1999.
    Effective date: April 12, 1999.
    Amendment Nos.: 234.
    Facility Operating License No. DPR-79: The amendment revises the 
TS.
    Date of initial notice in Federal Register: March 10, 1999 (64 FR 
11969). The supplemental letter of March 19, 1999 did not change the 
initial proposed no significant hazards condition determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 12, 1999.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

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

    Date of application for amendment: June 29, 1998, as supplemented 
by letter dated February 19, 1999.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.7.1.7 operability requirements to require four 
atmospheric steam dump (ASD) lines to be operable. Other changes were 
made to TS 3.7.1.7 to address action statements and surveillance 
requirements for the four ASD lines.
    Date of issuance: April 20, 1999.
    Effective date: April 20, 1999, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 131.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 9, 1998 (63 
FR 48271).
    The February 19, 1999, supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noticed, and did not change the staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 20, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Elmer Ellis Library, 
University of Missouri, Columbia, Missouri 65201.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of application for amendments: May 28, 1998, as supplemented 
December 11, 1998.
    Brief description of amendments: These amendments revise Technical 
Specifications (TS) to provide a specific numerical setting for reactor 
trip, reactor coolant pump trip, and auxiliary feedwater initiation on 
a loss of power to the 4 kilovolt (kV) buses. Changes to the bases for 
the affected TS sections are also being made.
    Date of issuance: April 23, 1999.
    Effective date: April 23, 1999.
    Amendment Nos.: Unit 1-189; Unit 2-194.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 15, 1998 (63 FR 
38208).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 23, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.

    Dated at Rockville, Maryland, this 28th day of April 1999.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 99-11119 Filed 5-4-99; 8:45 am]
BILLING CODE 7590-01-P