[Federal Register Volume 65, Number 76 (Wednesday, April 19, 2000)]
[Notices]
[Pages 21034-21046]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-9680]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 25, 2000, through April 7, 2000. The 
last biweekly notice was published on April 5, 2000 (65 FR 17908).

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.

[[Page 21035]]

    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications 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 May 19, 2000, 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 electronically from 
the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room). If a request for a hearing 
or petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, 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 electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 
and 2, Will County, Illinois

    Date of amendment request: January 20, 2000.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) to extend the allowable 
completion times for the required actions associated with restoration 
of an inoperable emergency diesel generator (EDG), and permit the 
performance of the 24-hour EDG endurance run during Modes 1 and 2 
(i.e., ``Power Operation'' or ``Startup''). A new requirement is 
proposed which

[[Page 21036]]

will require verification of the opposite unit's EDGs when the affected 
EDG is inoperable.
    Basis for proposed no significant hazards consideration 
determination. As required by 10 CFR 50.92(c), the staff's analysis of 
the issue of no significant hazards consideration is presented below:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes include the extension of the completion 
time for the EDGs from 72 hours to 14 days. In conjunction with the 
proposed change, a new required action is proposed to be 
incorporated into the TSs that will require verification of the 
operability of the opposite unit's EDGs while the affected EDG is 
inoperable. The proposed changes do not significantly increase the 
probability of occurrence of a previously evaluated accident because 
the EDGs are not initiators of accidents. Extending the completion 
times of the EDGs would not have any impact on the frequency of any 
accident previously evaluated and, therefore, the probability of a 
previously analyzed accident is unchanged. The proposed change to 
the completion time for EDGs will not result in any changes to the 
plant activities associated with EDG maintenance. The EDGs mitigate 
the consequences of previously evaluated accidents involving a loss 
of normal power, the safety-related buses and as such, the 
operability or availability of the EDGs could affect accident 
consequences. A configuration risk management program (CRMP) was 
developed and will be used to ensure that the risk impact of 
equipment out of service is appropriately evaluated prior to 
performing any maintenance activity. Increases in risk posed by 
potential combinations of equipment out of service during the EDG 
extended completion time will be managed under the CRMP. In 
addition, compensatory actions have been identified to mitigate an 
increase in risk. Procedures have been developed to implement the 
compensatory actions.
    The proposed changes also include a change to the TS 
surveillance requirement related to the conduct of the 24-hour EDG 
endurance run. Specifically, the change would permit the endurance 
run to be performed during Modes 1 and 2. The test configuration to 
be used is consistent with the configuration currently used during 
the one-hour monthly EDG tests currently conducted.
    The probability of an accident is not increased by performing 
the 24-hour endurance run in Modes 1 and 2 since the EDGs are used 
to support mitigation of the consequences of an accident. The 
failure of an EDG while testing is not an assumed initiator of a 
previously analyzed accident. The EDGs were designed to be tested by 
running in parallel with offsite power and design features such as 
protective devices were included. The proposed change does not 
affect parallel testing design features, the consequences of 
postulated failures during parallel testing, and postulated 
interactions with offsite power during parallel testing. If problems 
are encountered during testing, the EDG connection to the bus will 
be interrupted, allowing the offsite circuits to continue to supply 
the bus. Testing of the EDG does not affect the remainder of the 
safety-related equipment analyzed to mitigate the consequences of an 
accident. The control logic prevents potential damage of the 
emergency core cooling System (ECCS) equipment powered by the EDG to 
ensure that the ECCS equipment is available in the event of an 
actual safety injection with or without a Loss of Offsite Power 
(LOOP). Only one EDG per unit will be tested in parallel with the 
offsite sources at a time in order to prevent any grid disturbance 
from potentially affecting more than one EDG. Thus, during the test, 
the remaining EDG, which is capable of supplying power to mitigate 
all design basis accidents, will be available to respond normally to 
a start signal.
    To fully evaluate the effect of the proposed EDG TS changes, 
probabilistic risk assessment (PRA) methods and deterministic 
analyses were utilized. The results of the risk analysis show no 
significant increase in Core Damage Frequency (CDF) and Large Early 
Release Frequency (LERF).
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously analyzed.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not involve a physical change to the 
plant. No new equipment is being introduced, and installed equipment 
is not being operated in a new or different manner except for the 
following. The electrical lineup for performing the 24-hour run will 
be the same as the lineup for performance of the one-hour run, which 
is routinely performed at least once per month for each EDG. The 
difference between these two surveillances is in their duration. 
There is no change being made to the parameters within which the 
plant is operated. There are no setpoints affected by this proposed 
change at which protective or mitigative actions are initiated. This 
proposed changes will not alter the manner in which equipment 
operation is initiated, nor will the function demands on credited 
equipment be changed. No alteration in the procedures, which ensure 
that the plant remains within analyzed limits, is being proposed, 
and no change is being made to the procedures relied upon to respond 
to an off-normal event. As such, no new failure modes are being 
introduced. Other than the changes in duration of EDG unavailability 
from 72 hours to 14 days and on-line testing from 60 minutes to 24 
hours, the change does not alter assumptions made in the safety 
analysis and licensing basis.
    Therefore, these proposed changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed changes will extend the allowable Completion Times 
for the Required Actions associated with restoration of an 
inoperable EDG and allow the performance of the 24-hour endurance 
run at power. In conjunction with the proposed changes, a new 
required action is proposed to be incorporated into the TSs. The new 
action will require verification of the operability of the opposite 
unit's EDGs while the affected EDG is inoperable. These actions will 
be taken to ensure the availability of the remaining alternating 
current power sources to the affected engineered safety feature bus.
    The CRMP will be used to ensure that the risk impact of 
equipment out of service is appropriately evaluated prior to 
performing any maintenance activity. Increase in risk posed by 
potential combinations of equipment out of service during the EDG 
extended completion time will be managed under the CRMP. In 
addition, compensatory actions have been identified to mitigate 
increase in risk. Procedures have been developed to implement the 
compensatory actions.
    With regard to the proposed change for the 24-hour endurance 
run, the EDGs were designed to be tested by running in parallel with 
offsite power and, design features such as protective devices were 
included. The proposed change does not affect parallel testing 
design features, the consequences of postulated failures during 
parallel testing, and postulated interactions with offsite power 
during parallel testing. If problems are encountered during testing, 
the EDG connection to the bus will be interrupted allowing the 
offsite circuits to continue to supply the bus. Further, the EDG 
system design includes emergency override of the test mode for both 
accident conditions (safety injection) and loss of offsite power to 
permit a response to actual emergency signals and return control of 
the EDG to the automatic control system.
    Therefore, implementation of the proposed changes will not 
involve a significant reduction in the margin of safety.

    Based on the staff's analysis, 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.
    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.

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

    Date of amendment request: March 6, 2000.
    Description of amendment request: The proposed amendment would 
revise the Improved Technical Specification (ITS) Action Condition and 
Surveillance Requirement (SR) for the safety-related diesel-driven 
emergency feedwater pump (EFP-3). The ITS required

[[Page 21037]]

inventory volume for lube oil would be revised to agree with the actual 
test values and are included in the ITS Action Condition, SR and Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.

    1. Does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The revised lube oil requirements are being made to ensure EFP-3 
is capable of seven days of continuous operation. The proposed 
amendment provides the same functional requirement as previously 
approved. The EFW system is used for accident mitigation and is not 
an initiator of design basis accidents. Therefore, the probability 
of previously analyzed events is not affected by this change. No 
capability or design functions of EFP-3 or the emergency feedwater 
(EFW) system will change. The initial conditions for accidents that 
require EFW and accident mitigation capability of the EFW system 
will remain unchanged. Therefore, the proposed amendment will not 
increase the consequences of evaluated accidents.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The revised ITS Condition will ensure equipment is restored to 
operable status in accordance with previously approved timeframes 
and functional levels. The revised SR will assure the same 
functional requirement as the previously approved SR. Lube oil will 
be stored on-site and the lube oil inventory in the sump ensures 
adequate time to transfer the stored inventory into the engine. No 
new plant configurations or conditions are created by these revised 
ITS Conditions or SR. Therefore, the proposed amendment cannot 
create the possibility of an accident of a different type than 
previously evaluated in the Safety Analysis Report.
    3. Does not involve a significant reduction in the margin of 
safety.
    The proposed ITS Condition and SR changes ensure adequate lube 
oil inventory is available to operate EFP-3 for seven days. The 
proposed changes replace the calculated lube oil inventory values 
with a more conservative value derived from actual test data for 
EFP-3. The revised SR ensures the same functional requirement for a 
seven-day supply of lube oil for EFP-3 as was previously approved. 
Similarly, the revised ITS Condition ensures the same functional 
level as currently approved by requiring that a reduced lube oil 
inventory of less than seven days but more than six days is restored 
to the seven-day level within 48 hours. The revised SR allows the 
lube oil inventory to be stored off engine. The inventory in the 
EFP-3 sump and auxiliaries provides sufficient time to permit the 
transfer of stored inventory into the engine. EFP-3 is designed to 
allow monitoring of lube oil level and addition of lube oil while 
the engine is operating. Based on the above, the revised ITS meets 
the same intent as the currently approved specifications. Therefore, 
there is no reduction in the margin of safety associated with the 
proposed ITS amendment.

    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.
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC-A5A, P. O. Box 14042, St. Petersburg, Florida 
33733-4042.
    NRC Section Chief: Richard P. Correia.

Pacific Gas and Electric Company, Docket Nos. 50-275, Diablo Canyon 
Nuclear Power Plant, Unit No. 1, San Luis Obispo County, California

    Date of amendment requests: December 31, 1999, as supplemented by 
letter dated January 18, 2000.
    Description of amendment requests: The amendment would revise 
Section 2.C.(1) of Facility Operating License No. DPR-80 for the Diablo 
Canyon Power Plant, Unit No. 1 to authorize operation at reactor core 
power levels not to exceed 3411 megawatts thermal (100 percent rated 
power). This amendment would also (1) revise the definition in Section 
1.1 of the technical specifications (TS) of rated thermal power to 
reflect Unit 1 operation at the uprated reactor core power level, (2) 
change the reactor core safety limits in TS Figure 2.1.1-1 to reflect 
the current fuel type and provide additional margin for OTT 
and OPT setpoint calculations, and change the nominal full 
power Tavg in the OTT and OPT function in 
Notes 1 and 2 to TS Table 3.3.3-1.
    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.
    All previously evaluated accidents have been reviewed for the 
proposed increase in Unit 1 power rating, and these reviews are 
summarized in WCAP-14819, ``Pacific Gas and Electric Company Diablo 
Canyon Power Plant, Unit 1 3425 MWt [megawatt thermal] Uprating 
Program Licensing Report.'' The majority of the Diablo Canyon Power 
Plant (DCPP) accident analyses already bound the higher power rating 
of Unit 2 combined with the lower reactor coolant system (RCS) flow 
rate of Unit 1. Hence, the uprate has no impact on these previously 
evaluated accidents. This is also true of dose assessment, which 
remains based on the original 3568 MWt core source terms and is not 
impacted by the uprate.
    The previously evaluated accidents that are impacted by the 
uprate are large break loss-of-coolant accident (LOCA), small break 
LOCA, the OTT/ OPT setpoint calculations, and 
accidental depressurization of the RCS. The large break LOCA was 
reanalyzed for uprated conditions using best estimate methodology. 
The reanalysis demonstrated no increase in consequence and was 
approved by the NRC in License Amendments 121 (Unit 1) and 119 (Unit 
2). The small break LOCA was also reanalyzed, and continues to 
demonstrate a large margin to peak clad temperature limits. The 
current OTT/OPT setpoints are bounding for the 
Unit 1 uprated power conditions based on revising the reactor core 
safety limits in TS Figure 2.1.1-1 to credit the exclusive use of 
Vantage 5 fuel. The accidental RCS depressurization reanalysis shows 
that the departure from nucleate boiling ratio remains above the 
applicable limit value. In summary, no design or analysis acceptance 
criteria will be exceeded, the functional integrity of all plant 
systems are unaffected, and there is no impact on the integrity of 
the fission product barriers or assumed dose source terms. 
Therefore, the consequences of all previous evaluated accidents are 
not substantially increased.
    It was determined that there would be no impact on any component 
reliabilities assumed in the PRA model, and therefore no impact on 
the resulting core damage frequency. The PRA model envelopes both 
units, based on using the originally higher rated Unit 2 power 
level.
    The operation impacts of the proposed power increase were 
reviewed against the unit design capability, and it was determined 
that no system, structure, or component would exceed design 
conditions or loads. While the low pressure turbines see a small 
(less than 1.5 deg.F ) increase in temperature, the effect on 
missile generation probability is not significant. There is no 
significant increase in the probability of component failure, 
offsite power loss, or any other accident initiator. Therefore, the 
probability of all previously evaluated accidents is not 
substantially increased.
    Therefore, the proposed 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.
    Normal operation will not be substantially impacted by 
increasing the Unit 1 licensed power rating to match Unit 2. 
Procedures will be essentially unchanged, or where changes are 
required, they will be made to more closely resemble those in effect 
at Unit 2. Training will communicate all impacts to personnel and 
the plant simulator will be updated to match the power level of both

[[Page 21038]]

Units 1 and 2. There is, therefore, no possibility of a new or 
different kind of accident related to human performance.
    Plant systems, structures, and components have been evaluated 
for the proposed uprate. Most have identical counterparts in 
operation at Unit 2 at this higher power level. A few are slightly 
different, such as the generator cooling system, and for these the 
design margins have been reviewed and found to be acceptable. 
Therefore, there is no possibility of a new or different kind of 
accident related to system, structure, or component performance.
    Therefore, the proposed 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.
    The proposed changes do not involve a significant reduction in a 
margin of safety because the margin of safety associated with plant 
parameters as verified by the results of the accident analyses are 
within acceptable limits. As mentioned, most analyses demonstrating 
adequate margins of safety already assume the higher thermal power 
rating of Unit 2 and bound Unit 1 at the uprated thermal power 
conditions. The few transients that are reanalyzed meet the 
applicable acceptance criteria.
    The reactor core safety limits specified in TS Figure 2.1.1-1 
envelope operation with both 17x17 standard and 17x17 Vantage 5 
fuel. The proposed change revises the reactor core safety limits in 
Figure 2.1.1-1 to credit the exclusive use of Vantage 5 fuel. These 
revised safety limits will continue to satisfy fuel design criteria. 
The current OTT and OPT setpoints provide adequate 
margin to the revised reactor core safety limits at the uprated Unit 
1 conditions, which include a slightly higher nominal full power 
Tavg in Notes 1 and 2 to ITS Table 3.3.3-1.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

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: August 11, 1999.
    Description of amendment request: The proposed amendment clarifies 
the use of containment overpressure for ensuring adequate net positive 
suction head (NPSH) for the emergency core cooling system (ECCS) pumps.
    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 License Amendment Request does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    This proposed license amendment request does not involve any 
physical changes to plant Structures, Systems, or Components (SSC), 
or how the SSC are operated, maintained, and tested. The proposed 
changes involve the acceptability of taking credit for a specific 
amount of containment overpressure following the initiation of an 
event. This credit involves the mitigation of an event, and not 
prevention or identification of an event. Credit for containment 
overpressure is not considered a precursor to any event.
    Crediting containment overpressure does not turn an Anticipated 
Operational Occurrence (AOO) into an Abnormal Operational Transient 
(AOT) or a Design Basis Accident (DBA).
    Calculations performed in support of the license amendment 
request provide a conservative estimate of the Minimum Containment 
Pressure Available (MCPA) following all design and licensing basis 
events for which some amount of containment overpressure is 
required. The NPSH calculations for the Residual Heat Removal (RHR) 
and Core Spray (CS) pumps include conservative assumptions and input 
values ensure that, barring beyond-design-basis loss of containment 
integrity, adequate NPSH is provided to the RHR and CS pumps for the 
entire duration of any of these events.
    The proposed license amendment request makes a change to the 
PBAPS licensing basis to clearly define amount of containment 
overpressure allowed. This value is designated as the Containment 
Overpressure License (COPL). Conservative analyses have assured that 
the MCPA is always greater than this COPL for design basis events. 
Therefore, adequate NPSH is provided to the RHR and Core Spray pumps 
for all design and licensing basis events.
    The evaluation for MCPA and NPSH includes the consideration for 
any one single active failure. The worst-case single active failure 
is the failure of one electrical division. There is no credible 
single active failure that can compromise the containment integrity. 
The evaluation for MCPA and NPSH does not place any restrictions on 
system operation following a design or licensing basis event. The 
analysis concludes that adequate NPSH will be available, even 
assuming the worst single active failure.
    Therefore, the proposed license amendment request does not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    2. The proposed License Amendment Request does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    This proposed license amendment request does not involve any 
physical changes to plant SSC, or how the SSC are operated, 
maintained, and tested. This proposed license amendment request 
involves the acceptability of taking credit for some amount of 
containment overpressure following the initiation of an event. This 
credit involves the mitigation of an event, and not prevention or 
identification of an event. Credit for containment overpressure is 
not considered a precursor to any event. Worst-case single active 
failure (i.e., loss of one electrical division) was considered in 
the assessment of MCPA and COPR [containment overpressure required]. 
The supporting calculations indicate that adequate NPSH is provided 
to the RHR and CS pumps for all design and licensing basis events, 
even with the worst single active failure.
    Therefore, the proposed license amendment request does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed License Amendment Request does not involve a 
significant reduction in a margin of safety.
    The MCPA and NPSH analyses supporting this license amendment 
request include conservative assumptions and use conservative input 
values that are consistent with or bound the analytical limits of 
the PBAPS Technical Specifications. These analyses indicate that 
adequate NPSH margin is available for operation of the RHR and CS 
systems to meet their safety functions following any design or 
licensing basis event. This includes operation of RHR in Suppression 
Pool Cooling, Wetwell Spray, Drywell Spray, and Low Pressure Coolant 
Injection modes, and CS in Short Term and Long Term Spray Cooling. 
Therefore, the proposed license amendment request does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    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 California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of amendment requests: January 19, 2000 (PCN-512).
    Description of amendment requests: The amendment application 
proposes to revise the San Onofre Nuclear

[[Page 21039]]

Generating Station, Units 2 and 3, technical specifications (TSs) 
Surveillance Requirement (SR) 3.0.3.
    SR 3.0.3 allows compliance with the requirement to declare a 
limiting condition for operation not met to be delayed whenever it is 
discovered that a surveillance was not performed within its specified 
frequency (a missed surveillance). Presently, SR 3.0.3 allows a delay 
``up to 24 hours or up to the limit of the specified Frequency, 
whichever is less.'' The licensee proposes to revise the allowable 
delay ``up to 24 hours or up to the limit of the specified Frequency, 
whichever is greater.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of any accident previously evaluated?
    Response: No.
    The proposed change would extend the maximum allowable time for 
completing a Surveillance not performed within its specified 
Frequency (a missed Surveillance) without declaring the affected 
Limiting Condition For Operation (LCO) not met. The presently 
allowed time is up to 24 hours from the time of discovery or up to 
the limit of the specified Frequency, whichever is less. The 
proposed allowed time is up to 24 hours from the time of discovery 
or up to the limit of the specified Frequency, whichever is greater.
    Surveillances are rarely missed. This is demonstrated by a 
limited review of Licensee Event Reports (LERs), which found very 
few occurrences of missed Surveillances, given the number of LERs 
submitted and the large number of Surveillances performed. Moreover, 
Surveillances, whether performed inside or outside the required 
Frequency, nearly always verify conformance with Technical 
Specification requirements. This is demonstrated by a survey of 
selected licensees regarding entries into Surveillance Requirement 
(SR) 3.0.3. As stated in Generic Letter 87-09, ``* * * the vast 
majority of surveillances do in fact demonstrate that systems or 
components are operable.'' As stated in the SR 3.0.3 Bases, ``* * * 
the most probable result of any particular Surveillance being 
performed is the verification of conformance with the 
requirements.''
    Therefore, it is unlikely that plant equipment would be 
inoperable during the time period of up to 24 hours or up to the 
limit of the specified Frequency, whichever is greater, that would 
be allowed under the proposed change for the completion of a missed 
Surveillance.
    If, upon discovery of a missed Surveillance, it is known that 
the Surveillance would fail, SR 3.0.1 would require that the 
affected LCO be declared not met and the appropriate Condition(s) 
entered.
    Performance of some Surveillances carries with it a slight risk, 
either from making some plant equipment temporarily inoperable or 
from performing plant manipulations, or both. The increase in plant 
risk from performing such Surveillances, combined with the 
confidence that a Surveillance test will be satisfactory when 
performed, together provide justification for extending the current 
allowable time to up to 24 hours or up to the specified Frequency, 
whichever is greater.
    The foregoing discussion demonstrates that the probability or 
consequences of any accident previously evaluated will not be 
significantly increased by the proposed change.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    This amendment request is administrative in nature and does not 
involve any change to plant equipment. Therefore, it will not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    This amendment request does not change the manner in which 
safety limits or limiting safety settings are determined.
    As discussed above, Surveillances are rarely missed, and, when 
performed, Surveillances nearly always verify conformance with 
Technical Specification requirements, making it unlikely that plant 
equipment would be inoperable during the time period of up to 24 
hours or up to the limit of the specified Frequency, whichever is 
greater, that would be allowed under the proposed change for the 
completion of a missed Surveillance.
    Therefore, the proposed change 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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

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 17, 2000.
    Description of amendment request: Revise Technical Specification 3/
4.7.4 to revise the surveillance requirements (SRs) 4.7.4.b.1 and 
4.7.4.b.2 to incorporate the wording from the Westinghouse Standard 
Improved Technical Specifications (NUREG-1431) and to delete SR 
4.7.4.b.3. SR 4.7.4.b.3 requires verifying at least once per 18 months 
that each screen wash booster pump and the traveling screen start 
automatically on a safety injection test signal. The licensee also 
proposed changes to the Technical Specifications Bases associated with 
the Technical Specification changes and administrative changes to the 
Bases Index.
    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 proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    NUREG 1431 related changes:
    Incorporating the NUREG 1431 [Westinghouse Standard Improved 
Technical Specifications] wording for SR 4.7.4.b.1 and SR 4.7.4.b.2 
does not significantly increase the probability of an accident 
because the surveillance testing of the Essential Cooling Water 
system has no effect on accident initiation probability. This change 
does not significantly increase the consequences of an accident 
because the surveillance requirements still provide adequate 
assurance that the Essential Cooling Water system can provide its 
design function.
    Screen wash system changes:
    Eliminating the requirement for the Essential Cooling Water 
traveling screens and screen wash booster pumps to start on a safety 
injection signal does not increase the probability of any accident 
previously evaluated. The traveling screens and the screen wash 
booster pumps have no potential for initiating an accident. 
Eliminating the requirement for the traveling screens and the screen 
wash booster pumps to start on a safety injection signal does not 
increase the consequences of any accident previously evaluated. A 
control system is provided to automatically start and stop the 
traveling screens during normal operation. A high differential water 
level sensed across any traveling screen alarms in the control room 
and automatically starts the screen wash booster pump and, after 
reaching adequate screen wash pressure, starts the traveling screen. 
A safety injection signal is not needed for this function. In 
addition, there are no circumstances associated with any event 
requiring a safety injection signal that would cause a high 
differential water level across the traveling screen.
    The changes to the Bases Index are administrative and have no 
relevance to accident probability or consequences.

[[Page 21040]]

    Based on the above, STPNOC [STP Nuclear Operating Company] 
concludes that the proposed change does not increase the probability 
or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    NUREG 1431 related changes:
    Incorporation of the NUREG 1431 wording into the surveillance 
requirements does not create the possibility of a new or different 
kind of accident because the surveillance requirements are not 
substantially changed and do not involve any different operational 
configurations for the station.
    Screen wash system changes:
    Elimination of the requirement to start the traveling screen and 
screen wash booster pump on a safety injection signal will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated. As discussed above, the traveling 
screens and screen wash booster pump have no potential to initiate 
an accident. In addition, STPNOC is not proposing any different 
operational configurations for the station.
    The changes to the Bases Index are administrative and have no 
relevance to accidents.
    Based on the above, STPNOC concludes that the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    NUREG 1431 related changes:
    Incorporation of the NUREG 1431 wording for SR 4.7.4.b.1 and SR 
4.7.4.b.2 does not significantly change the way the surveillance 
requirements will be performed. The Surveillance Requirements still 
provide adequate assurance that the Essential Cooling Water will 
perform its function. There is no change in the operational 
configuration of the plant. Consequently, the changes to these 
surveillance requirements do not significantly affect the margin of 
safety.
    Screen wash system changes:
    Elimination of the requirement for the traveling screen and 
screen wash booster pump to start on a safety injection signal will 
not prevent the traveling screen and screen wash booster pump to 
start when required. The systems will start automatically without 
the need for a safety injection signal. In addition, there is no 
design basis or mechanistic reason to postulate the need to 
automatically start the traveling screens or screen wash booster 
pump on a safety injection signal.
    The changes to the Bases Index are administrative and have no 
relevance to the safety margin.
    Based on the above, STPNOC concludes that the proposed change 
does not involve a significant decrease 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.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: March 31, 2000 (ET 00-0018).
    Description of amendment request: The proposed amendment would 
modify the actions for Limiting Condition for Operation (LCO) 3.7.9, 
``Ultimate Heat Sink (UHS),'' of the technical specifications (TSs). 
The proposed new Action A would allow the plant to operate with the 
plant inlet water temperature of the UHS above 90  deg.F, if the 
licensee verified the required cooling capacity within 4 hours and once 
per 12 hours thereafter, but that the plant would be shut down if the 
water temperature exceeded 94  deg.F. This would change the current 
requirement to shut down the plant if the inlet water temperature of 
the UHS exceeded the 90  deg.F. The time to shut down the plant is not 
being changed in the amendment request.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not involve any physical alteration of 
plant systems, structures or components. The proposed change 
provides an allowance for the plant to continue operation with [the] 
plant inlet water temperature [of the UHS] in excess of the current 
Technical Specification limit of 90  deg.F with the verification 
that required cooling capacity being maintained and [the plant inlet 
water] temperature  94  deg.F. The 94  deg.F limit is 
less than the design limit of 95  deg.F for associated plant 
components. The plant inlet water temperature is not assumed to be 
an initiating condition of any accident analysis evaluated in the 
Updated Safety Analysis Report (USAR). Therefore, the allowance for 
the [plant inlet] water temperature to be in excess of the current 
limit does not involve an increase in the probability of an accident 
previously evaluated in the USAR. The UHS supports OPERABILITY of 
safety related systems used to mitigate the consequences of an 
accident. Plant operation for brief periods with [the] plant inlet 
water temperature greater than 90  deg.F up to 94  deg.F will not 
adversely affect the OPERABILITY of these safety related systems and 
will not adversely impact the ability of these systems to perform 
their safety related functions. Therefore, the proposed change does 
not involve a significant increase in the consequences of an 
accident previously evaluated in the USAR.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve any physical alteration of 
plant systems, structures or components. The temperature of the 
plant inlet water being greater than 90  deg.F but less than or 
equal to 94  deg.F (with the main cooling lake dam intact) does not 
introduce new failure mechanisms for systems, structures or 
components not already considered in the USAR. The 94  deg.F limit 
is less than the design limit of 95  deg.F for associated plant 
components. Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated is not created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change will allow an increase in [the] plant inlet 
water temperature above the current Technical Specification limit of 
90  deg.F for the UHS, provided [the] UHS temperature is maintained 
below 95  deg.F and that the required cooling capacity is verified 
maintained within 4 hours and once per 12 hours thereafter. 
Additionally, the plant inlet water temperature will be verified to 
be 94  deg.F once per 12 hours. The proposed change does 
not alter any safety limits, limiting safety system settings, or 
limiting conditions for operation, and the proposed changes provide 
continued assurance that with a plant inlet water temperature > 90 
deg.F, the design temperature of safety related equipment are 
maintained within acceptable limits such that a safe shutdown of the 
plant can be performed. In addition, avoiding a plant transient 
during environmental conditions that could challenge the stability 
of the Electrical Power System offsets any perceptible reduction in 
the margin of safety as a result of the proposed change. Thus, the 
proposed change does not involve a significant reduction in any 
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.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Section Chief: Stephen Dembek.

[[Page 21041]]

Previously Published Notices of Consideration of Issuance of 
Amendments to Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, and Opportunity for a Hearing

    The following notice was previously published as a separate 
individual notice. The notice content was the same as above. It was 
published as an individual notice either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. It is repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

    Date of application for amendments: March 29, 2000 (TS-402).
    Brief description of amendments: Changes Technical Specification 3/
4.6.4.1 ``Secondary Containment'' to permit maintenance on a secondary 
containment access door when one or more units are operating and the 
other door is closed.
    Date of publication of individual notice in the Federal Register: 
April 6, 2000 (65 FR 18141)
    Expiration date of individual notice: April 20, 2000.

Notice of Issuance of Amendments to Facility Operating Licenses

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

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of application for amendment: August 26, 1999, as supplemented 
on February 24, and March 14, 2000.
    Brief description of amendment: This amendment revises Technical 
Specification 3/4.9.4, and its associated bases, to allow the personnel 
airlock and certain other containment penetrations to remain open 
during refueling operations provided specific administrative controls 
are met. This amendment is approved for use during refueling outage 9 
and operating cycle 10.
    Date of issuance: March 27, 2000.
    Effective date: March 27, 2000.
    Amendment No.: 97.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in  Federal Register: October 6, 1999 (64 FR 
54374).
    The February 24, and March 14, 2000, submittals contained 
clarifying information only, and did not change the initial no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated March 27, 2000.
    No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of application for amendments: May 20, 1999.
    Brief description of amendments: The amendments changed the 
Technical Specification (TS) value for the minimum suppression chamber 
water level to a more conservative value.
    Date of issuance: March 30, 2000.
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 176 & 172.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46426).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 30, 2000.
    No significant hazards consideration comments received: No.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of application for amendment: December 17, 1999, as 
supplemented March 8, 2000.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 2.1, ``Safety Limits (SLs),'' changing the safety 
limit minimum critical power ratio limits in TS 2.1.1.2 to reflect the 
results of cycle-specific calculations performed for Fermi 2 operating 
Cycle 8.
    Date of issuance: March 30, 2000.
    Effective date: As of the date of issuance and shall be implemented 
prior to the startup from the seventh refueling outage.
    Amendment No.: 138.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4269).
    The March 8, 2000, letter provided clarifying information that 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 March 30, 2000.
    No significant hazards consideration comments received: No.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of application for amendment: July 30, 1999, as supplemented 
December 17, 1999, and March 1, 2000.

[[Page 21042]]

    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.3.1.1, ``Reactor Protection System (RPS) 
Instrumentation,'' to reflect the activation of the automatic trip 
associated with the oscillation power range monitor (OPRM). The 
amendment also revises TS 3.4.1, ``Recirculation Loops Operating,'' to 
remove requirements related to the manual detection and suppression of 
core thermal-hydraulic instabilities because these actions are no 
longer necessary after the OPRM upscale function is activated.
    Date of issuance: March 31, 2000.
    Effective date: As of the date of issuance and shall be implemented 
prior to the startup from the seventh refueling outage.
    Amendment No.: 139.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 3, 1999 (64 FR 
59800).
    The December 17, 1999, and March 1, 2000, letters provided 
clarifying information that 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 March 31, 2000.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: August 6, 1998, as supplemented by 
letter dated February 16, 2000.
    Brief description of amendment: The amendment revises the minimum 
and the maximum concentration limits for the sodium hydroxide tank. The 
amendment also deletes the maximum specified tank volume and revises 
the minimum specified tank volume to refer to the parameter used in the 
safety analysis with no allowance for instrument uncertainty.
    Date of issuance: March 28, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 206.
    Facility Operating License No. DPR-51: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 10, 1999 (64 
FR 6695).
    The February 16, 2000, letter provided clarifying information that 
did not change the scope of the August 6, 1998, application and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 28, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 27, 2000.
    Brief description of amendment: The amendment deleted the current 
requirements of Technical Specification (TS) 4.7.9.1.2.d, ``Source 
installed in the Boronometer,'' associated with the installed 
boronometer sealed source. The source was recently removed and stored, 
and the requirements of TS 4.7.9.1.2.d are no longer applicable.
    Date of issuance: March 24, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 212.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 23, 2000 (65 
FR 9007).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 24, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 27, 2000.
    Brief description of amendment: The amendment relocated the 
schedule for the withdrawal of reactor vessel material surveillance 
specimens, from the Technical Specifications to the Safety Analysis 
Report, pursuant to the guidance provided in Generic Letter 91-01, 
``Removal of the Schedule for the Withdrawal of Reactor Vessel Material 
Specimens From Technical Specifications.'' Changes to the related Bases 
were also made. In addition, the proposed change to the surveillance 
specimen removal schedule was approved.
    Date of issuance: April 4, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 213.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 23, 2000 (65 
FR 9007).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 4, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 26, 1999, as supplemented 
by submittal dated December 7, 1999.
    Brief description of amendment: This amendment permits 
implementation of 10 CFR Part 50, Appendix J, Option B, and reference 
Regulatory Guide 1.163, ``Performance-Based Containment Leak Test 
Program,'' dated September 1995, which specifies a method acceptable to 
the NRC for complying with Option B. These changes relate only to Type 
B and C (local) leakage rate testing. The use of Option B for Type A 
(integrated) leakage rate testing was approved on February 22, 1996, by 
License Amendment No. 205.
    Date of issuance: March 28, 2000.
    Effective date: Immediately as of its date of issuance and shall be 
implemented within 120 days.
    Amendment No.: 240.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46437).
    The letter of December 7, 1999, contained clarifying information 
and did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 28, 2000.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio

    Date of application for amendment: September 9, 1999, as 
supplemented by submittal dated February 28, 2000.
    Brief description of amendment: This amendment includes nine minor, 
unrelated revisions to the technical specifications (TSs). These 
revisions, which are minor in both content and safety significance, 
include

[[Page 21043]]

clarifications and editorial changes to the TSs.
    Date of issuance: March 30, 2000.
    Effective date: Immediately as of the date of issuance and shall be 
implemented within 90 days.
    Amendment No.: 111.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in  Federal Register: November 3, 1999 (64 
FR 59803).
    The supplemental information contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 30, 2000.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendments: December 3, 1998.
    Brief description of amendments: The amendments made administrative 
changes to several Technical Specifications to remove obsolete 
information, provide consistency between Unit 1 and Unit 2, provide 
consistency with the Standard Technical Specifications, provide 
clarification, and correct typographical errors.
    Date of issuance: March 31, 2000.
    Effective date: March 31, 2000, with full implementation within 30 
days.
    Amendment Nos.: 243 and 224.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1999 (64 FR 
47535).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 31, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 6, 1999, as supplemented March 
17, 2000.
    Brief description of amendment: Amendment to technical 
specifications changes the safety limit minimum critical power ratio 
(SLMCPR) from 1.06 to 1.08 for two recirculation loop operation and 
from 1.07 to 1.09 for single recirculation loop operation.
    Date of issuance: March 31, 2000.
    Effective date: March 31, 2000, to be implemented within 30 days.
    Amendment No.: 182.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73093).
    The March 17, 2000, letter provided additional clarifying 
information that was within the scope of the original application and 
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 March 31, 2000.
    No significant hazards consideration comments received: No.

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 6, 1999, as 
supplemented by letters dated February 22 and March 14, 2000.
    Brief description of amendment: The amendment modifies the 
Technical Specification (TS) surveillance requirements associated with 
ensuring a limited number of charging and high pressure safety 
injection pumps are incapable of injecting into the Reactor Coolant 
System when the plant is shutdown. In addition, the TS Bases are 
modified to address these changes.
    Date of issuance: March 30, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 243.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4285).
    The February 22 and March 14, 2000, supplemental letters provided 
clarifying information that did not change the staff's original no 
significant hazards consideration determination or expand the scope of 
the application as published.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 30, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 7, 1999.
    Brief description of amendment: The amendment removes the current 
special exception which precludes applying the 18-month functional 
testing surveillance to the Steam Generator Hydraulic Snubbers for 
Technical Specification 3/4.7.8, ``Plant Systems, Snubbers.''
    Date of issuance: March 31, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 244.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in  Federal Register: January 26, 2000 (65 
FR 4283).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 31, 2000.
    No significant hazards consideration comments received: No.

Portland General Electric Company, et al., Docket No. 50-344, Trojan 
Nuclear Plant, Columbia County, Oregon

    Date of application for amendment: January 29, 1998.
    Brief description of amendment: The amendment deletes paragraph 2.D 
of Facility Operating License No. NPF-1 and revises the Permanently 
Defueled Technical Specifications (PDTS) by deleting PDTS 5.7.1.1(b). 
These changes remove the requirements for a security program at the 10 
CFR part 50 licensed site once the spent nuclear fuel has been 
relocated to the 10 CFR part 72 licensed Independent Spent Fuel Storage 
Installation.
    Date of issuance: April 6, 2000.
    Effective date: April 6, 2000, to be implemented within 30 days 
after the transfer of the last cask of spent nuclear fuel from the 
spent fuel pool to the independent spent fuel storage installation is 
complete.
    Amendment No.: 203.
    Facility Operating License No. NPF-1: The amendment changes the 
Operating License and the Permanently Defueled Technical 
Specifications.
    Date of initial notice in Federal Register: September 8, 1999 (64 
FR 48865).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 6, 2000.
    No significant hazards consideration comments received: No.

[[Page 21044]]

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of application for amendments: July 23, 1999, as supplemented 
September 13, 1999, and January 31, 2000.
    Brief description of amendments: The amendments revised the 
Technical Specifications to remove the restriction on performing the 
24-hour endurance run test of emergency diesel generators (EDGs) every 
18 months only during shutdown. Additionally, for Salem Unit 1 only, a 
note associated with a one-time extension of a surveillance requirement 
was deleted.
    Date of issuance: March 30, 2000.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 229 and 210.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in  Federal Register: October 6, 1999 (64 FR 
54380).
    The January 31, 2000, supplement provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination and did not expand the scope of the 
original application as published.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 30, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 4, 1999.
    Brief description of amendments: The amendments revised the 
Technical Specifications 5.5.6, ``Prestressed Concrete Containment 
Tendon Surveillance Program,'' to incorporate three exceptions to 
Regulatory Guide (RG) 1.35, Revision 2, 1976. The exceptions concern 
the number of tendons detensioned, inspection of concrete adjacent to 
vertical tendons, and the time during which areas adjacent to tendons 
are inspected.
    Date of issuance: March 27, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1-112; Unit 2-90.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in  Federal Register: February 9, 2000 (65 
FR 6411).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 27, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 29, 1999 (TS 99-04).
    Brief description of amendments: The amendments delete Sequoyah 
Nuclear Plant Technical Specification (TS) monthly surveillance test on 
the auxiliary feedwater suction pressure switches.
    Date of issuance: March 29, 2000.
    Effective date: As of the date of issuance, to be implemented no 
later than 45 days after issuance.
    Amendment Nos.: 253 and 244.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TS.
    Date of initial notice in  Federal Register: May 19, 1999 (64 FR 
27325).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 29, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 30, 1999.
    Brief description of amendments: Requirements related to 
containment isolation valves that were located in two different 
sections of the technical specifications were consolidated into one 
section. Also, conditions relating to or usage of a check valve as an 
isolation device was clarified.
    Date of issuance: March 29, 2000.
    Effective date: March 29, 2000.
    Amendment Nos.: 254 and 245.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in  Federal Register: October 6, 1999 (64 FR 
54382).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 29, 2000.
    No significant hazards consideration comments received: No.

TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: January 13, 2000.
    Brief description of amendments: The amendments: (1) Revise 
Technical Specification 3.8.3 (Condition B and Surveillance Requirement 
(SR) 3.8.3.2) to increase the required emergency diesel generator (EDG) 
lube oil inventory values; (2) Revise SR 3.8.3.2, for EDG lube oil 
inventory, to add a note stating that the surveillance is not required 
to be performed until the diesel has been in shutdown greater than 10 
hours; and (3) Delete the footnote associated with SR 3.8.4.7 which 
provided a ``one time only'' alternative to battery testing 
requirements.
    Date of issuance: March 24, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 75 and 75.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 23, 2000 (65 
FR 9012). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 24, 2000.
    No significant hazards consideration comments received: No

TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: January 13, 2000.
    Brief description of amendments: The amendments add ``NOTE 3'' to 
Surveillance Requirement 3.3.1.10 to allow entry into MODES 2 or 1 
without the performance of N-16 detector plateau verification until 72 
hours after achieving equilibrium conditions at greater than or equal 
to 90 percent rated thermal power.
    Date of issuance: March 24, 2000.
    Effective date:E As of the date of issuance and shall be 
implemented within 30 days from the date of issuance.
    Amendment Nos.: 76 and 76.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 23, 2000 (65 
FR 9013)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 24, 2000.
    No significant hazards consideration comments received: No.

[[Page 21045]]

TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: January 13, 2000.
    Brief description of amendments: The amendments add ``NOTE 3'' to 
Surveillance Requirement 3.3.1.10 to allow entry into MODES 2 or 1 
without the performance of N-16 detector plateau verification until 72 
hours after achieving equilibrium conditions at greater than or equal 
to 90 percent rated thermal power.
    Date of issuance: March 24, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendments Nos.: 76 and 76.
    Facility Operating License Nos. NPF-84 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 23, 2000 (65 
FR 9013).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 24, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 3, 1999 (ULNRC-04158).
    Brief description of amendment: The amendment changed Section 
5.6.6, ``Reactor Coolant System (RCS) Pressure and Temperature Limits 
Report (PTLR),'' of the improved Technical Specifications (ITS) that 
were issued on May 28, 1999, in Amendment No. 133. The current 
Technical Specifications (CTS) remain in effect until the ITS are 
implemented on or before April 30, 2000. The changes to the ITS approve 
the use of the PTLR by the licensee to make changes to the plant 
pressure temperature limits and low temperature over pressure 
protection limits without prior NRC staff approval, in accordance with 
Generic Letter 96-03, ``Relocation of the Pressure Temperature Limit 
Curves and Low Temperature Overpressure Protection System Limits,'' 
dated January 31, 1996. The changes (1) add the word criticality to ITS 
subsection 5.6.6.a as one of the reactor conditions for which RCS 
pressure and temperature limits will be determined, (2) add the phrase 
``and COMS PORV,'' where COMS PORV stands for cold overpressure 
mitigation system power operated relief valve, to the introductory 
paragraph of ITS subsection 5.6.6.b to show that the analytical methods 
listed in the subsection are also the COMS PORV, and (3) replace the 
two documents listed in ITS subsection 5.6.6.b by the reference to the 
NRC letter that approves use of the PTLR and the Westinghouse topical 
report, WCAP-14040-NP-A, Revision 2, ``Methodology Used to Develop Cold 
Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown 
Limit Curves,'' dated January 1996, that provides the methodology that 
will be used by licensee in using the PTLR report. The current plant 
pressure temperature limits and low temperature overpressure protection 
limits are in the CTS and were approved in Amendment No. 124, which was 
issued April 2, 1998.
    Date of issuance: March 24, 2000.
    Effective date: March 24, 2000, to be implemented no later than 
April 30, 2000.
    Amendment No.: 134.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73101).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 24, 2000.
    No significant hazards consideration comments received: No

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

    Date of application for amendment: January 14, 2000, as 
supplemented by letter dated February 17, 2000 (ULNRC-04172 and -
04187).
    Brief description of amendment: The amendment revised several 
sections of the improved Technical Specification (ITSs) to correct 14 
editorial errors made in either (1) the application dated May 15, 1997, 
(and supplementary letters) for the ITSs, or (2) the certified copy of 
the ITSs that was submitted in the licensee's letters of May 27 and 28, 
1999. The ITSs were issued as Amendment No. 133 by the staff in its 
letter of May 28, 1999, and will be implemented by the licensee to 
replace the current TSs by April 30, 2000.
    Date of issuance: March 27, 2000.
    Effective date: March 27, 2000, to be implemented by April 30, 
2000.
    Amendment No.: 135.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 23, 2000 (65 
FR 9013) and February 25, 2000 (65 FR 10118).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 27, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 15, 1999, as supplemented 
on January 14, 2000.
    Brief description of amendment: The amendment revises Technical 
Specifications (TSs) Sections 3.1/4.1 Reactor Protection System and 
3.2/4.2 Protective Instrument Systems instrumentation, tables, and the 
associated Bases to increase the surveillance test intervals (STIs), 
add allowable out-of-service times (AOTs), replace generic emergency 
core cooling system actions for inoperable instrument channels with 
function-specific actions, and relocate selected trip functions from 
the TSs to a Vermont Yankee controlled document. In addition, revision 
to TS Section 3.1/4.1 Reactor Protection System and the associated 
Bases is proposed to remove the RUN Mode APRM Downscale/IRM High Flux/
Inoperative Scram Trip Function (APRM Downscale RUN Mode SCRAM). The 
submittal also proposes to implement editorial corrections and 
administrative changes that do not alter the meaning or intent of the 
requirements.
    Date of Issuance: April 3, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 186.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 20, 1999 (64 FR 
56535).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated April 3, 2000.
    No significant hazards consideration comments received: No

    Dated at Rockville, Maryland, this 12th day of April 2000.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-9680 Filed 4-18-00; 8:45 am]
BILLING CODE 7590-01-P