[Federal Register Volume 65, Number 86 (Wednesday, May 3, 2000)]
[Notices]
[Pages 25761-25775]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-10743]



[[Page 25761]]

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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 April 8, 2000, through April 21, 2000. The 
last biweekly notice was published on April 19, 2000 (65 FR 21034).

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 
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 June 2, 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

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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).

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 amendment request: April 7, 2000.
    Description of amendment request: The proposed amendment would 
revise Harris Nuclear Plant (HNP) Technical Specification (TS) 3/4.7.6, 
``Control Room Emergency Filtration System,'' TS 3/4.7.7, ``Reactor 
Auxiliary Building Emergency Exhaust System,'' TS 3/4.9.12, ``Fuel 
Handling Building Emergency Exhaust System,'' and the associated Bases. 
Specifically, the licensee proposes to revise these TS to provide an 
Action when the Control Room Emergency Filtration System or Reactor 
Auxiliary Building Emergency Exhaust System ventilation boundary is 
inoperable and a note that allows an applicable ventilation boundary to 
be open intermittently under administrative controls. Additionally, the 
licensee proposes to modify TS 3/4.3.3.1, ``Radiation Monitoring for 
Plant Operations,'' to provide consistency between the applicability of 
the Control Room Emergency Filtration System and the radiation monitors 
that initiate a Control Room Isolation signal.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Ventilation systems are not accident initiating systems as 
described in the Final Safety Analysis Report. The changes are based 
on the low probability of a design basis accident occurring during 
the 24 hour completion time and compensatory measures available to 
minimize dose consequences of an event during this time. The 
proposed change does not affect another Structure, System, or 
Component.
    Current HNP TS do not restrict fuel movement in the fuel 
handling or loads over spent fuel pools concurrent with an 
inoperable Control Room Emergency Filtration System. Providing 
restrictions for fuel movement and loads over spent fuel pools 
preserves assumptions made in the fuel handling accident analysis. 
The addition of applicability requirements for fuel movement and 
movement of loads over spent fuel pools is consistent with NUREG-
1431, Revision 1, and is more restrictive than current HNP TS.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Ventilation systems are not accident initiating systems as 
described in the Final Safety Analysis Report. As such, the failure 
of the ventilation system to operate properly or a premature 
actuation of the ventilation system can not initiate an accident.
    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 amendment does not involve a significant 
reduction in the margin of safety.
    The proposed change to ventilation systems does not 
significantly affect any of the parameters that relate to the margin 
of safety as described in the Bases of the TS or the FSAR [Final 
Safety Analysis Report]. Accordingly, NRC Acceptance Limits are not 
affected by this change. The changes are based on the low 
probability of a design basis accident occurring during the 24 hour 
completion time and compensatory measures available to minimize dose 
consequences of an event during this time.
    The addition of applicability requirements for Control Room 
Emergency Filtration System during movement of irradiated fuel 
assemblies and movement loads over spent fuel pools provide 
additional margin not currently provided in HNP TS.
    Therefore, the proposed 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 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: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Richard P. Correia.

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 amendment request: April 12, 2000.
    Description of amendment request: The proposed amendment would 
revise Harris Nuclear Plant (HNP) Technical Specification (TS) 3/
4.4.9.2, ``Pressure/Temperature (P-T) Limits--Reactor Coolant System,'' 
and TS 3/4.4.9.4, ``Overpressure Protection System,'' and the 
associated Bases. Specifically, the licensee proposes to revise the 
applicable TS to incorporate results of the Reactor Vessel Surveillance 
Program capsule analysis. A summary report was previously submitted to 
the NRC (HNP-99-157, dated 11/9/99) in accordance

[[Page 25763]]

with Title 10 of the Code of Federal Regulations, Part 50 (10 CFR 50), 
Appendix H. Additionally, the licensee's submittal requested an 
exemption to 10 CFR 50.60 (a), based on American Society of Mechanical 
Engineers (ASME) Code Case N-640 and WCAP-15315. The exemption request 
will be evaluated separate from the proposed license 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:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes affect operations of the Reactor Coolant 
System (RCS) components when the RCS temperature is below 350 deg. 
F. The revisions to P-T limits and allowable heatup and cooldown 
rate limits are consistent with ASME code cases which have been 
authorized for other licensees by the NRC. The proposed changes 
modify the setpoint of the pressurizer PORVS [power operated relief 
valves] for LTOPS [low temperature overpressure setpoints]. Changes 
to the LTOPS setpoints applicable below 350 deg. F effectively 
increase the allowable operating pressure for any given temperature 
during shutdown. These changes do not result in conditions which are 
outside of the design basis for RCS Structures, Systems, and 
Components (SSCs). Therefore, the proposed changes do not alter the 
characteristics of the RCS SSCs adversely, and therefore do not 
impact the performance of the RCS SSCs during power operations.
    The revised P-T limits and heatup and cooldown rate limits are 
within the design capabilities of the RCS SSCs and pressure control 
systems. While the proposed new P-T limits are less restrictive than 
the current Technical [Specification] requirements, they assure that 
plant operation is within the design capacity of the reactor vessel 
materials. Therefore, the RCS capability as a fission product 
barrier is not compromised.
    The changes to the LTOPS setpoints do not affect accident 
consequences since no credit is assumed for operation of LTOPS to 
mitigate accidents.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve new plant components or 
procedures, but only revise existing operational limits and 
setpoints. These changes do not place SSCs in conditions outside of 
their design basis, and the revised operating setpoints and 
conditions are within the capability of the plant control systems.
    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 amendment does not involve a significant 
reduction in the margin of safety.
    The proposed changes to the P-T limits and LTOPS setpoints 
change the calculational method from that described in the bases to 
one based on ASME Code Case N-640, and on WCAP-15315. The effect of 
this change is to allow plant operation with different limits, but 
still with adequate margins to assure the integrity of the reactor 
vessel and RCS.
    Therefore, the proposed 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 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: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: March 15, 2000.
    Description of amendment request: The proposed amendments would 
revise the ultimate heat sink temperature in the technical 
specifications from 98 deg.F to 100 deg.F.
    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 any accident previously evaluated?
    Analyzed accidents are assumed to be initiated by the failure of 
plant structures, systems or components. An inoperable Ultimate Heat 
Sink (UHS), which is the source of water for the Essential Service 
Water (ESW) System, is not considered as an initiator of any 
analyzed events. The analyses for Braidwood Station, Units 1 and 2, 
assume a UHS temperature of 100 deg.F. Therefore, continued 
operation with a UHS temperature less than or equal to 100 deg.F 
will not increase the probability of occurrence of any accident 
previously evaluated in the Updated Final Safety Analysis Report 
(UFSAR). The proposed change does not involve any physical 
alteration of plant systems, structures or components. A UHS 
temperature of up to 100 deg.F does not increase the failure rate of 
systems, structures or components because the systems, structures or 
components are rated and analyzed for operation with ESW 
temperatures of 100 deg.F and the design allows for higher 
temperatures than at which they presently operate.
    The basis provided in Regulatory Guide 1.27 ``Ultimate Heat Sink 
for Nuclear Power Plants,'' Revision 2, dated January 1976, was 
employed for the temperature analysis of the Braidwood Station UHS 
to implement General Design Criteria (GDC) 44, ``Cooling water,'' 
and GDC 2, ``Design bases for protection against natural 
phenomena,'' of Appendix A to 10 CFR Part 50. This Regulatory Guide 
was employed for both the original design/licensing basis of the 
Braidwood Station UHS and a subsequent evaluation which investigated 
the potential for increasing the average water temperature of the 
UHS from 98 deg.F to 100 deg.F. The heat loads 
selected for the UHS analysis considered one Braidwood Station unit 
in a Loss of Coolant Accident (LOCA) condition concurrent with a 
Loss Of Offsite Power (LOOP) event and the remaining Braidwood 
Station unit undergoing a safe non-accident shutdown. In the 
analysis, these heat loads are removed by the UHS using only ESW 
pumps. The main cooling pond is conservatively assumed not to be 
available at the start of the event. The analysis shows that with an 
initial UHS temperature of 100 deg.F, the required heat loads can be 
met for 30 days while maintaining ESW temperatures at acceptable 
values.
    Based on the above, it has been demonstrated that the operation 
at an initial UHS temperature of 100 deg. F at the start 
of the design basis event will result in the continued ability of 
the equipment and components supplied by the ESW system to perform 
their intended safety functions.
    Therefore, increasing the average water temperature limit of the 
UHS from 98 deg. F to 100 deg. F does not 
increase the consequences of any accident previously evaluated. 
Raising this limit does not introduce any new equipment, equipment 
modifications, or any new or different modes of plant operation, nor 
does it affect the operational characteristics of any equipment or 
systems. Therefore, this proposed change does not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.
    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 a physical alteration of 
the units. There is no change being made to the parameters within 
which the units are operated that is not bounded by the analyses. 
There are no setpoints at which protective or mitigative actions are 
initiated that are affected by this proposed change. This proposed 
change 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 that ensure the units 
remain within analyzed limits is proposed, and no change is being 
made to

[[Page 25764]]

procedures relied upon to respond to an off-normal event. As such, 
no new failure modes are being introduced. The proposed change does 
not alter assumptions made in the safety analysis.
    Increasing the allowed average water temperature of the UHS in 
Technical Specification (TS) 3.7.9, ``Ultimate Heat Sink (UHS),'' 
has no impact on plant operation. Operating at the proposed higher 
temperature limit does not introduce new failure mechanisms for 
systems, structures or components. The engineering analyses 
performed to support the change to UHS temperature limit provides 
the basis to conclude that the equipment is designed for operation 
at elevated temperatures. The current analyses and calculations 
assume a UHS temperature of 100 deg. F, which is within the design 
limits of the affected equipment. In addition, design and 
construction codes applied to the affected structures, systems and 
components provided sufficient margin to accommodate the proposed 
temperature change.
    Therefore, this proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change allows operation with the UHS temperature 
100 deg. F. The margin of safety is determined by the 
design and qualification of the plant equipment, the operation of 
the plant within analyzed limits, and the point at which protective 
or mitigative actions are initiated. The proposed change does not 
impact these factors. The existing analyses already assume an 
initial UHS temperature of 100 deg. F for design basis accident 
conditions. There are no required design changes or equipment 
performance parameter changes associated with this change. No 
protection setpoints are affected as a result of this change. This 
temperature increase has been confirmed to not change the 
operational characteristics of the design of any equipment or 
system. All accident analysis assumptions and conditions will 
continue to be met. Thus, the proposed increase in UHS temperature 
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 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.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690≤980767.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: February 11, 1999 (Reference NRC-00-
0023).
    Description of amendment request: The proposed amendment will 
revise the Technical Specifications by: (1) Deleting Specification A.8, 
the definition of ``Primary System'' which will no longer be necessary 
if the specifications related to the Primary System cover gas system 
are deleted; (2) deleting Specification D, which specifies the 
requirements for the Primary System cover gas system; (3) deleting the 
portion of Specification H.1 that specifies the surveillance 
requirements for the Primary System pressure alarms; (4) deleting Table 
H.1 item a, the Primary System pressure alarm points; (5) deleting 
Specification H.3.b, the requirement to perform surveillances of the 
door and seals around the machinery dome; (6) deleting Specification 
I.7.b, which requires procedures for maintaining cover gas supply; and 
(7) deleting Specification I.9.d, which requires keeping records of 
CO2 cover gas usage. The above-listed changes would allow 
the licensee to remove the Primary System cover gas system from 
service, an action that would allow the licensee to begin work on 
removing the remaining residual sodium from the Primary System. The 
licensee also requested an editorial change in Table H.1 item b.1, to 
change ``Bldg.'' to ``Building''.
    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 using the standards in 10 CFR 50.92(c). The licensee's 
analysis is presented below:

    (1) The proposed change does not involve a significant increase 
in the probability or consequences of an accident.
    Removing the cover gas from the [P]rimary [S]ystem, opening the 
[P]rimary [S]ystem, and cleaning out sodium residues will not 
significantly increase the probability of an accident occurring as 
long as the probability of an uncontrolled water reaction with the 
sodium is not significantly increased. This is done by conducting 
cutting operations and sodium reactions under control conditions. 
Removing the cover gas or opening the system will not take place 
until the current asbestos abatement project in the Reactor Building 
is complete since water is being used. The abatement is expected to 
be completed this winter before the license amendment will be 
approved. Note that EPA approval for dry removal has been obtained 
for where there is a risk of water coming into contact with sodium. 
The successful dismantling of the secondary sodium system piping in 
the Steam Generator building demonstrates that sodium systems can be 
cut open safely. The sodium residue processing in the secondary 
sodium storage tanks demonstrates sodium cleanup can be conducted 
safely. The consequences of an accident will not be increased by 
removing the cover gas, opening the [P]rimary [S]ystem, or reacting 
the sodium residues because the previously analyzed accidents 
already involve the release of all the radioactive material in the 
[P]rimary [S]ystem and all the radioactive material in the liquid 
waste system. The maximum postulated dose to the public was analyzed 
to be within the 10 CFR [Part] 20 limit of 100 mrem/year. This 
change will not increase the amount of radioactive material 
available to be released.
    (2) The proposed change does not create the possibility of a new 
or different accident from any previously evaluated.
    Removing the cover gas from the [P]rimary [S]ystem, opening the 
[P]rimary [S]ystem, and cleaning out the sodium residues will not 
create a new or different type of accident. A sodium accident has 
been previously evaluated. The only other type of accident which 
could possibly be caused by removing the [P]rimary [S]ystem cover 
gas, opening the [P]rimary [S]ystem, or processing primary sodium 
residues is a liquid waste release, which is highly unlikely. A 
liquid waste accident has also been previously evaluated. Only the 
[P]rimary [S]ystem and other equipment or piping containing primary 
sodium is expected to be affected by this change.
    (3) The proposed change does not involve a significant reduction 
in a margin of safety.
    Only a relatively small amount of sodium remains in the 
[P]rimary [S]ystem and other equipment containing primary sodium. 
Some of this residual may have been converted to sodium carbonate, 
leaving even less sodium remaining. The cover gas was a good 
precaution, especially for systems sitting unattended for many 
years. It prevented moisture from intruding into the systems and 
reacting with the sodium residues. It prevented oxygen from entering 
and reacting with any hydrogen formed from reactions of water and 
sodium. Discontinuing the use of cover gas slightly reduces the 
margin of safety, but not significantly. Removing the cover gas does 
not, in itself, introduce water into the system in an uncontrolled 
manner. Even if slight amounts of moisture from humidity in the air 
enter over the next year or two until the sodium is removed while 
the system is opened or unsealed, the system volume is large enough 
that the system will be able to dissipate any small reactions that 
occur. In addition, the calculated consequence[s] of releasing the 
radioactive material in the primary sodium is small and well within 
10 CFR [Part] 20 and Technical Specification limits.
    The planned processing of sodium residues is evaluated as 
releasing the radioactive material to the atmosphere, as planned 
release using controls specified in the Technical Specifications for 
gaseous effluents. For these reasons, the proposed change does not 
involve a significant reduction in the margin of safety.


[[Page 25765]]


    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, NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration. 
Attorney for licensee: John Flynn, Esquire, Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Branch Chief: Larry W. Camper.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, 
Turkey Point Plant, Units 3 and 4, Dade County, Florida

    Date of amendment request: November 30, 1999, as supplemented March 
8, 2000.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications to allow the use of credit for 
soluble boron in the spent fuel pool criticality analyses. In addition, 
a revised criticality analysis for the fresh fuel storage racks will be 
used to update the licensing bases. Criticality analyses were performed 
using the methodology developed by the Westinghouse Owners Group and 
described in WCAP-14416-NP-A, Revision 1, Westinghouse Spent Fuel Rack 
Criticality Analysis Methodology.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    There is no increase in the probability of a fuel assembly drop 
accident in the Spent Fuel Pool (SFP) when considering the presence 
of soluble boron in the SFP water for criticality control. The 
handling of the fuel assemblies in the SFP has always been performed 
in borated water. The consequences of a fuel assembly drop accident 
in the SFP are not affected when considering the presence of soluble 
boron.
    There is no increase in the probability of the accidental 
misloading of spent fuel assemblies into the SFP racks when 
considering the presence of soluble boron in the pool water for 
criticality control. Fuel assembly placement will continue to be 
controlled pursuant to approved fuel handling procedures and will be 
in accordance with the Technical Specification (TS) spent fuel rack 
storage limitations. There is no increase in the consequences of the 
accidental misloading of spent fuel assemblies into the SFP racks 
because criticality analyses demonstrate that the pool will remain 
subcritical following an accidental misloading if the pool contains 
an adequate boron concentration. The proposed TS ensure that an 
adequate SFP boron concentration will be maintained. There is no 
increase in the probability of the loss of normal cooling to the SFP 
water when considering the presence of soluble boron in the pool 
water for subcriticality control since a high concentration of 
soluble boron has always been maintained in the SFP water.
    A loss of normal cooling to the SFP water causes an increase in 
the temperature of the water passing through the stored fuel 
assemblies. This causes a decrease in water density, which would 
result in a net increase in reactivity when soluble boron is present 
in the water and Boraflex neutron absorber panels are present in the 
racks. However, the additional negative reactivity provided by the 
1950 ppm boron concentration limit, above that provided by the 
concentration required (650 ppm) to maintain Keff less 
than or equal to 0.95, will compensate for the increased reactivity 
which could result from a loss of SFP cooling event. Because 
adequate soluble boron will be maintained in the SFP water, the 
consequences of a loss of normal cooling to the SFP will not be 
increased.
    The Fresh Fuel racks are analyzed by employing the 
``Westinghouse Spent Fuel Rack Criticality Analysis Methodology'' 
approved by the NRC and described in WCAP-14416, NP-A, Revision 1. 
Only the method for Fresh Fuel storage racks criticality 
calculations has changed. The method of handling fuel, the maximum 
fuel enrichment, and the limiting values for criticality have not 
changed. Therefore, there is no change in the margin of safety for 
the Fresh Fuel storage racks.
    Therefore, based on the conclusions of the above analysis, the 
proposed changes will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    Spent fuel handling accidents are not new or different types of 
accidents, they have been analyzed in Section 14.2.1 of the Updated 
Final Safety Analysis Report (UFSAR). Criticality accidents in the 
SFP are not new or different types of accidents, they have been 
analyzed in the UFSAR and in the spent fuel storage criticality 
analysis. Current TS 3/4.9.14 already contains a limit on the SFP 
boron concentration. The boron concentration in the SFP has always 
been maintained near the limit of the RWST boron concentration for 
refueling purposes. The current TS boron concentration requirement 
for the SFP water conservatively bounds the boration assumptions of 
the revised criticality analyses. Since soluble boron has always 
been maintained in the SFP water, the implementation of this 
requirement for criticality purposes will have no effect on normal 
pool operations and maintenance.
    Since soluble boron has always been present in the SFP, a 
dilution of the SFP soluble boron has always been a possibility. 
However, it was shown in the SFP dilution analysis that a dilution 
of the Turkey Point SFP which could increase the spent fuel storage 
rack Keff to greater than 0.95 is not a credible event. 
Therefore, the implementation of limitations on the SFP boron 
concentration for criticality purposes will not result in the 
possibility of a new or different kind of accident.
    Proposed TS 3/4.9.14 Table 3.9-1 specifies the requirements for 
the spent fuel rack storage, which is currently contained in the TS. 
These proposed new SFP storage limitations are consistent with the 
assumptions made in the spent fuel rack criticality analysis, and 
will not have any significant effect on normal SFP operations and 
maintenance, and will not create any possibility of a new or 
different kind of accident. Verifications will continue to be 
performed to ensure that the SFP loading configuration meets 
specified requirements.
    The Fresh Fuel racks are analyzed by employing the 
``Westinghouse Spent Fuel Rack Criticality Analysis Methodology'' 
approved by the NRC and described in WCAP-14416, NP-A, Revision 1. 
Only the method for Fresh Fuel storage racks criticality 
calculations has changed. The method of handling fuel, the maximum 
fuel enrichment, and the limiting values for criticality have not 
changed. Therefore, there is no change in the margin of safety for 
the Fresh Fuel storage racks.
    As discussed above, the proposed changes will not create the 
possibility of a new or different kind of accident from any 
previously evaluated. There is no significant change in plant 
configuration, equipment design or equipment.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed TS changes will provide adequate safety margin to 
ensure that the stored fuel assembly array will always remain 
subcritical. Those limits are based on a plant specific criticality 
analyses performed in accordance with the NRC approved Westinghouse 
Spent Fuel Rack criticality analysis methodology.
    The criticality analysis takes credit for soluble boron to 
ensure that Keff will be less than or equal to 0.95 under 
normal circumstances. Storage configurations have been defined using 
a 95/95 Keff calculation to ensure that the spent fuel 
rack Keff will be less than 1.0 with no soluble boron. 
Soluble boron credit is used to provide safety margin by maintaining 
Keff less than or equal to 0.95, including uncertainties, 
tolerances, and accident conditions in the presence of SFP soluble 
boron.
    The loss of substantial amounts of soluble boron from the SFP 
that could lead to exceeding a Keff of 0.95 has been 
evaluated in the SFP Dilution analysis and shown to be not credible.
    The analysis shows that the dilution of the SFP boron 
concentration from 1950 ppm to 650 ppm is not credible. When this 
result is combined with the results from the 95/95 criticality 
analyses, which show that the spent fuel rack Keff will 
remain less than 1.0 when flooded with unborated water, it provides 
a level of safety comparable to the

[[Page 25766]]

conservative criticality analysis methodology required by ANSI 57.2-
1983, NUREG-0800, and Regulatory Guide 1.13.
    The Fresh Fuel racks are analyzed by employing the 
``Westinghouse Spent Fuel Rack Criticality Analysis Methodology'' 
approved by the NRC and described in WCAP-14416, NP-A, Revision 1. 
Only the method for Fresh Fuel storage racks criticality 
calculations has changed. The method of handling fuel, the maximum 
fuel enrichment, and the limiting values for criticality have not 
changed. Therefore, there is no change in the margin of safety for 
the Fresh Fuel storage racks.
    Therefore, the proposed changes in these license amendments will 
not result in a significant reduction in the plant's 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.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: June 3, 1999, as supplemented on 
December 22, 1999
    Description of amendment request: The proposed amendment would 
permit continued plant operation with a maximum of two inoperable 
recirculation loops, provided certain conditions are met. Oyster 
Creek's Technical Specifications (TSs), Section 3.3.F.2 currently 
permit operation with 4 of the 5 recirculation loops with certain 
constraints. If only 3 loops are operable, however, the TSs require 
plant shutdown within 12 hours. Analysis indicates that the plant may 
be safely operated at 90 percent power with three operable 
recirculation loops.
    Two definitions are added to Section 1 of the TSs to specify the 
difference between an idle recirculation loop and an isolated 
recirculation loop. These definitions have been incorporated into the 
specification to provide an explicit description of acceptable valve 
configurations. In addition, several paragraphs have been added to the 
Bases of Section 3.3 and one paragraph in the Bases of Section 3.10 has 
been modified. In each case the Bases section has been segmented from 
the specification, which affects the pagination of the 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. 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 TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    When operating with two inoperable recirculation loops, the 
proposed section 3.3.F.2.b requires that the reactor core thermal 
power not exceed 90% of rated power. This is a physical limitation 
of the plant conditions because maximum power is about 90% of rated 
power at the maximum recirculation flow with only three 
recirculation pumps operating. As such, the 90% of rated power 
becomes a limiting condition for three-loop operation. The licensee 
states that the results of this analysis conform to all the 
requirements of 10 CFR 50.46 and Appendix K.
    The licensee analyzed recirculation pump trip transients for 
single and multiple pump trips. Although the transient in general is 
very mild, the licensee considers the case of simultaneous trip of 
all five pumps to be the limiting event among all possible 
recirculation pump trip events. For three-loop operation, given the 
requirement that the power level be maintained at or below 90% of 
rated power, the transient resulting from the loss of all three 
pumps would be bounded by the five-pump-trip event.
    The proposed change, which permits three loop operation with a 
maximum of two idle or one idle and one fully isolated loop, will 
provide adequate safety margins during transient and accident 
conditions. The proposed changes do not affect any accident 
precursors because the accident occurrence is not dependent on the 
number of operating recirculation loops. Therefore, the probability 
of an accident previously evaluated is not increased. The proposed 
TS change will assure the ability of systems to perform their 
intended function. Therefore, the proposed changes will not 
introduce a significant increase in the consequences of an accident 
previously evaluated. Therefore, the probability of occurrence or 
the consequences of an accident previously evaluated in the Safety 
Analysis Report (SAR) will not increase as a result of these 
changes.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes permit three-loop operation with a maximum 
of two idle or one idle and one fully isolated loop. The licensee 
considers the case of simultaneous trip of all the five pumps to be 
the limiting event among all possible recirculation pump trip 
events. For three-loop operation, given the requirement that the 
power level be maintained at or below 90% of rated power, the 
transient resulting from the loss of all three pumps would be 
bounded by the five-pump-trip event.
    The proposed changes will not create a possibility for an 
accident or transient of a different type than any previously 
identified in the SAR.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The proposed changes will not decrease the margin of safety as 
defined in the basis of any Technical Specification. All relevant 
transient and accident scenarios have been analyzed for the 
conditions of three-loop operation and have demonstrated adequate 
margin to safety limits. Therefore, the proposed changes do not 
involve a significant reduction in the margin of safety. They 
neither adversely affect the performance characteristics of systems 
nor do they affect the ability of systems to perform their intended 
function. Therefore, the proposed changes do not involve a 
significant reduction in the 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.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: M. Gamberoni, Acting.

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

    Date of amendment request: March 17, 2000.
    Description of amendment request: This amendment request proposes 
to revise the Cooper Nuclear Station Technical Specifications to 
incorporate the recommended Generic Letter 99-02, ``Laboratory Testing 
of Nuclear-Grade Activated Charcoal,'' laboratory testing protocol of 
American Society for Testing and Materials (ASTM) D3803-1989 for 
Engineered Safety Feature ventilation system charcoal samples.
    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 proposed charcoal testing changes and explicit reference to 
American Society for Testing and Materials (ASTM) D3803-1989 
nuclear-grade activated charcoal test protocol do not affect 
Engineered Safety Feature (ESF) ventilation system operation or 
performance, reliability, actuation setpoints, or accident 
mitigation capabilities. The proposed

[[Page 25767]]

changes also do not affect the operation and performance of any 
other equipment important to safety at Cooper Nuclear Station (CNS). 
ASTM D3803-1989 is a more accurate and demanding test which ensures 
that the charcoal filter efficiencies assumed in the CNS accident 
dose analysis are maintained. The proposed changes involve ESF 
ventilation system charcoal testing only and do not affect accident 
initiators. Therefore the proposed changes do not significantly 
increase the probability or consequences of an accident previously 
evaluated, as revised by the design basis accident radiological 
assessment calculational methodology revision submitted to the NRC 
under Reference 3 [in the March 17, 2000, amendment request].
    2. Does not create the possibility for a new or different kind 
of accident from any accident previously evaluated.
    The charcoal testing changes, and explicit reference to ASTM 
D3803-1989 nuclear-grade activated charcoal test protocol, do not 
affect ESF ventilation system operation or performance, or the 
operation and performance of any other equipment important to safety 
at CNS. The proposed changes clarify and explicitly identify the 
testing of the ESF ventilation system charcoal samples. No new or 
different accident scenarios, transient precursors, failure 
mechanisms, plant operating modes, or limiting single failures are 
introduced as a result of these changes. Therefore, the possibility 
of a new or different kind of accident from that previously 
evaluated, as revised by the design basis accident radiological 
assessment calculational methodology revision submitted to the NRC 
under Reference 3, is not created by this change.
    3. Does not create a significant reduction in the margin of 
safety.
    The required performance of the ESF ventilation systems 
following a design basis accident is not impacted by utilizing a 
more demanding protocol for charcoal testing. Thus, the margin of 
safety assumed in the CNS accident analysis, as revised by the 
design basis accident radiological assessment calculational 
methodology revision submitted to the NRC under Reference 3, is 
maintained. Revising the Technical Specifications to clarify 
charcoal testing methodology and explicitly referencing the charcoal 
[adsorber] testing being performed does not affect ESF ventilation 
system performance or operation, or the operation and performance of 
any other equipment important to safety at CNS. Therefore, these 
changes do not result in 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: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis 
Obispo County, California

    Date of amendment requests: December 23, 1999.
    Description of amendment requests: The proposed amendment would 
revise improved TS (ITS) 5.5.9.d.1.j)(iv) to change the tube support 
plate (TSP) intersections that are excluded from application of steam 
generator (SG) tube voltage based repair criteria for outside diameter 
stress corrosion cracking indications at TSPs.
    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.
    Application of a smaller wedge region exclusion zone [due to 
loss of coolant accident (LOCA) plus safe shutdown earthquake (SSE)] 
and a new seventh tube support plate (TSP) bending stress exclusion 
zone (due to feedline break (FLB)/steamline break (SLB) plus SSE) 
with respect to alternate repair criteria (ARC), does not increase 
the probability of tube burst or leakage following a postulated main 
steam line break (MSLB). Exclusion zones tubes will be inspected by 
bobbin every outage and by rotating pancake coil (RPC) if bobbin 
detects degradation. Tubes containing RPC-confirmed crack-like 
degradation at wedge region exclusion zone intersections and at the 
seventh TSP bending exclusion zone intersections will be plugged.
    Tube burst criteria are inherently satisfied during normal 
operating conditions because of the proximity of the TSP. It is 
conservatively assumed that the entire crevice region is uncovered 
because of TSP displacement during the secondary side blowdown of a 
MSLB. Therefore, during a postulated MSLB accident, tube burst 
capability must exceed the Regulatory Guide 1.121 criterion 
requiring a margin of 1.43 times the SLB pressure differential on 
tube burst.
    Relative to the expected leakage during accident condition 
loadings, a postulated MSLB outside of containment, but upstream of 
the main steam isolation valve, represents the most limiting 
radiological condition. The steam generator (SG) tubes are subjected 
to an increase in differential pressure following a MSLB, resulting 
in a postulated increase in leakage and associated offsite doses. 
Leakage following a MSLB bypasses containment.
    Following each inspection, condition monitoring will be 
performed to verify that tube burst and leakage performance criteria 
were satisfied for all degradation.
    Therefore, the proposed changes do 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.
    Implementation of revised ARC exclusion zones does not introduce 
any significant change to the plant design basis. Use of new 
exclusion zones does not create a mechanism which could result in an 
accident in the free span. It is expected that for all plant 
conditions, neither a single nor multiple tube rupture event would 
likely occur in a SG where ARC exclusion zones have been applied.
    Therefore, the proposed changes do 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.
    Revised wedge region exclusion zones are based on a DCPP-
specific analysis under locked tube conditions for the combined 
effects of a LOCA and SSE. The number of wedge region tubes that are 
predicted to collapse has been decreased when compared to the prior 
analysis, which used highly conservative assumptions. The revised 
analysis incorporates DCPP-specific LOCA and seismic loads that were 
not available when the prior analysis was performed. However, the 
revised analysis also yields conservative results, such that the 
number of tubes in the exclusion zone (244 per SG) bound the number 
of tubes calculated to collapse (144 per SG). Tubes located in the 
revised wedge region exclusion zone will continue to be subject to 
enhanced eddy current inspection requirements and will be excluded 
from application of ARC. Thus, existing tube integrity requirements 
apply to these tubes and the margin of safety is not reduced.
    New seventh TSP bending exclusion zones are also based on a 
DCPP-specific analysis under locked tube conditions for the combined 
effects of a FLB/SLB and SSE. The analysis yields conservative 
results, such that 914 tubes per SG at the seventh TSP are assumed 
to exceed the Westinghouse lower tolerance limit yield stress of the 
tubing. Tubes located in the seventh TSP bending exclusion zone will 
be subject to enhanced eddy current inspection requirements and will 
be excluded from application of ARC. Thus, existing tube integrity 
requirements apply to these tubes and the margin of safety is not 
reduced.
    Therefore, the 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 requests involve no significant hazards consideration.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric

[[Page 25768]]

Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: April 6, 2000.
    Description of amendment request: The Virgil C. Summer Nuclear 
Station Technical Specifications are being revised to change the 
definitions and surveillance requirements for response time testing of 
the Engineered Safety Feature Actuation System (ESFAS) and the Reactor 
Trip System (RTS). These changes will permit the verification of 
response time, whereas the current definitions imply the response time 
must be measured.
    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?
    This change to the Technical specifications (TS) does not result 
in a condition where the design, material, and construction 
standards that were applicable prior to the change are altered. The 
same RTS and ESFAS instrumentation is being used; the time response 
allocations/modeling assumptions in the Final Safety Analysis Report 
(FSAR) Chapter 15 analyses are still the same; only the method of 
verifying the time response is changed. The proposed change will not 
modify any system interface and could not increase the likelihood of 
an accident since these events are independent of this change. The 
proposed change will not change, degrade or prevent actions or alter 
any assumptions previously made in evaluating the radiological 
consequences of an accident described in the FSAR.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This change does not alter the performance of process protection 
racks, Nuclear Instrumentation, and logic systems used in the plant 
protection systems. These systems will still have response time 
verified by test before being placed in operational service. 
Changing the method of periodically verifying instrument[ation] for 
these systems (assuring equipment operability) from response time 
testing to calibration and channel checks will not create any new 
accident initiators or scenarios. Periodic surveillance of these 
systems will continue and may be used to detect degradation that 
could cause the response time to exceed the total allowance. The 
total time response allowance for each function bounds all 
degradation that cannot be detected by periodic surveillance. 
Implementation of the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does this change involve a significant reduction in margin of 
safety?
    This change does not affect the total system response time 
assumed in the safety analysis. The periodic system response time 
verification method for the process protection racks, Nuclear 
Instrumentation, and logic systems is modified to allow the use of 
actual test data or engineering data. The method of verification 
still provides assurance that the total system response is within 
that defined in the safety analysis, since calibration tests will 
continue to be performed and may be used to detect any degradation 
which might cause the system response time to exceed the total 
allowance. The total response time allowance for each function 
bounds all degradation that cannot be detected by periodic 
surveillance. Based on the above, it is concluded that the proposed 
change does not result in a significant reduction in margin with 
respect to plant safety.
    Pursuant to 10 CFR 50.91, the preceding analyses provides a 
determination that the proposed Technical Specifications change 
poses no significant hazard as delineated by 10 CFR 50.92.

    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: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: April 6, 2000.
    Description of amendment request: The proposed Technical 
Specifications change request (TSCR) seeks to remove the prescriptive 
testing requirements of TS 4.8.1.1.2.i.2 to allow the ASME Code Class 3 
portions of the diesel fuel oil system to be pressure tested in 
accordance with Section XI of the ASME Boiler and Pressure Vessel Code 
as required by TS 4.0.5. This will permit the use of Code Case N-498-1 
as accepted by Regulatory Guide 1.147, Revision 12, for assessment of 
the diesel fuel oil system pressure boundary integrity.
    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?
    Industry experience has shown that an inservice leak test 
conducted at normal operating temperature and pressure is just as 
effective at finding leakage as a hydrostatic test conducted at 110% 
of the design pressure.
    Therefore, there is no increase in the probability or 
consequences of previously evaluated accidents.
    Also note that the Diesel Generator Fuel Oil System is not 
specifically modeled in the VCSNS Probability Risk Assessment. It is 
contained in the diesel generator fail to run event that has a 
probability of 5.8E-2. If the diesel generator fuel oil system had 
been modeled, pipe ruptures would not have been included because 
they would be dominated by failure of other components such as check 
valves which have failure probabilities several orders of magnitude 
higher.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed TSCR provides an alternative method of leak 
detection for the required 10-year inservice inspection. It does not 
result in an operational condition different from that which has 
already been considered by TS. Therefore, the change does not create 
the possibility of a new or different kind of accident or 
malfunction.
    3. Does this change involve a significant reduction in margin of 
safety?
    The alternative method of leak detection has no impact on the 
consequences of any analyzed accident and does not significantly 
change the failure probability of equipment which provides 
protection for the health and safety of the public. Therefore, there 
is no significant decrease in the margin of safety.
    Pursuant to 10 CFR 50.91, the preceding analyses provides a 
determination that the proposed Technical Specifications change 
poses no significant hazard as delineated by 10 CFR 50.92.

    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: Randolph R. Mahan, South Carolina Electric & 
Gas

[[Page 25769]]

Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: Richard L. Emch, Jr.

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: March 30, 2000 (PCN-515).
    Description of amendment requests: The amendment application 
proposes to revise the San Onofre Nuclear Generating Station, Units 2 
and 3, Technical Specification (TS) 3.6.6.1, ``Containment Spray and 
Cooling Systems,'' and the associated Bases. The proposed change would 
revise the Allowed Outage Time (AOT) for a single inoperable train of 
the containment spray system from 72 hours to 7 days and revise the 
combined AOT of 10 days which appears in both Conditions A and C of 
Limiting Condition for Operation 3.6.6.1 from 10 days to 14 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    This proposed change is a request to revise Technical 
Specification 3.6.6.1, ``Containment Spray and Cooling Systems'' and 
the associated Bases. The proposed change revises the Allowed Outage 
Time (AOT) for a single inoperable train of the Containment Spray 
System (CSS) from 72 hours to 7 days. The following changes are 
proposed for the Containment Spray System as described in Technical 
Specification (TS) 3.6.6.1:
    a. The Allowed Outage Time (AOT) for a single train of 
Containment Spray (Condition A of LCO 3.6.6.1) is extended from 72 
hours to 7 days.
    b. The Combined AOT of 10 days which appears in both Conditions 
A and C of LCO 3.6.6.1 is extended from 10 days to 14 days.
    c. The Bases of TS 3.6.6.1 are revised to reflect the changes 
described above.
    The Containment Spray System is an Engineered Safety Feature 
(ESF) system. Inoperable Containment Spray components are not 
considered to be accident initiators. Therefore, this change does 
not involve an increase in the probability of an accident previously 
evaluated.
    The proposed AOT for the Containment Spray System does impact 
the ability to mitigate accident sequences. Therefore, to fully 
evaluate the effects of the proposed CSS AOT extension, 
Probabilistic Safety Analysis (PSA) methods were utilized. The 
results of these analyses show no significant increase in core 
damage frequency. As a result, there would be no significant 
increase in the consequences of an accident previously evaluated. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    (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 proposed change does not change the design, configuration, 
or method of operation of the plant.
    Therefore, this proposed change will not create the possibility 
of a new or different kind of accident from any accident that has 
been 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
    The proposed change does not affect the limiting conditions for 
operation or their bases that are used in the deterministic analyses 
to establish the margin of safety. PSA evaluations were used to 
evaluate these changes.
    Therefore, there will be no significant reduction in a margin of 
safety as a result of 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 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.

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: March 17, 2000.
    Description of amendment request: The proposed changes would modify 
the voltage setting limits specified in Technical Specification (TS) 
Table 3.7-4, page 3.7-26, item 7 for the emergency bus degraded 
voltage, and revise the loss of voltage setpoints from a percentage of 
nominal bus voltage to an actual bus voltage value. The degraded 
voltage setting limit is being changed to increase the minimum 
allowable bus voltage to improve long-term motor performance in the 
event of operation with bus voltage less than nominal. The emergency 
bus loss of voltage setting limit is being revised to better address 
expected relay performance over time (i.e., setting drift). Section 
3.6.B, page 3.6-1, of the TS would be changed to revise the required 
reactor coolant system conditions from the existing wording of ``350 
degrees F or 450 psig'' to ``350 degrees F and 450 psig.''
    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:

    We have reviewed the proposed change against the criteria of 10 
CFR 50.92 and have concluded that the change does not pose a 
significant safety hazards consideration as defined therein. 
Specifically, operation of Surry Power Station with the proposed 
change will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    No increase in the probability of occurrence or consequences of 
an accident previously evaluated will result from the proposed 
change in the setting limits for the emergency bus degraded voltage 
and loss of voltage relay setpoints. The proposed change only 
affects actuation limits and therefore has no bearing on the 
probability of an accident. Neither the logic nor the function of 
the undervoltage protection circuits is being changed, nor is 
circuit or equipment reliability being reduced. The higher degraded 
voltage relay setpoint limit will improve motor terminal voltage, 
and thus promote longer motor life. Changing the setpoint limit for 
the loss of voltage relays will better characterize the relays' 
capabilities and facilitate calibration. Further, the performance 
characteristics of the electrical distribution system and components 
supplied (motors, etc.) are not being altered, and compliance with 
GDC-17 [General Design Criterion] is being maintained. The 
electrical distribution system remains capable of performing its 
safety function without spurious separation of the emergency buses 
from offsite power. If offsite power is lost, the capability of the 
EDG's [emergency diesel generators] to perform their safety function 
is not altered. Therefore, the probability of an accident previously 
evaluated is not increased.
    The consequences of an accident do not increase since the 
proposed change implements setting limits that will continue to 
ensure that adequate voltages will be available for the continuous 
operation of safety-related equipment required to function to 
mitigate a design basis accident. The proposed setting limits for 
the emergency bus degraded voltage and loss of voltage bound the 
setpoints and initial conditions assumed in the accident analyses 
and ensure that appropriate protection is maintained.
    The editorial change is administrative in nature and 
consequently does not affect the probability or consequences of an 
accident in any way.

[[Page 25770]]

    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Implementing the proposed Technical Specifications emergency bus 
degraded voltage and loss of voltage relay setting limits cannot 
create the possibility of a new or different kind of accident than 
any accident previously evaluated. Revising the setpoint setting 
limits does not introduce any new accident precursors, and operation 
of the electrical distribution system and the undervoltage relaying 
schemes is unchanged. Raising the setting limit for emergency bus 
degraded voltage and decreasing the setting limit for emergency bus 
loss of voltage do not introduce any new accident precursors or 
modes of operation. The relays will continue to detect undervoltage 
conditions and transfer safety loads to the emergency diesel 
generators at a voltage level adequate to ensure proper safety 
equipment performance and to prevent long-term equipment degradation 
due to undervoltage conditions. The proposed setting limits include 
adequate tolerances to calibrate the undervoltage relays while 
ensuring that emergency bus voltages remain above analytical limits. 
As noted above, the performance characteristics of the electrical 
distribution system and the components being supplied are not being 
altered, and compliance with GDC-17 is being maintained. The 
proposed Technical Specifications change will ensure that 
appropriate electrical protection is available as assumed in the 
safety analysis.
    The editorial change is administrative in nature and 
consequently 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.
    The proposed change continues to ensure that adequate voltage is 
available for safety-related equipment relied upon to respond to a 
design basis accident. The proposed setting limit for degraded bus 
voltage is conservative with respect to the existing Technical 
Specifications and ensures an adequate safety margin is being 
maintained. Further, the setting limit is maintained low enough to 
prevent spurious actuations given expected offsite grid voltages. 
The setting limit for the emergency bus loss of voltage relays is 
being changed to better characterize the relays' capabilities and to 
facilitate calibration. While the loss of bus voltage setting limit 
is being reduced, sustained bus voltage in this range is not 
credible. Furthermore, there is no safety limit associated with the 
loss of voltage setting limit.
    The proposed change continues to ensure that the setting limits 
for the emergency bus degraded voltage and loss of voltage relays 
bound the setpoints and initial conditions assumed in the accident 
analyses and ensures that appropriate electrical protection is 
maintained. The editorial change is administrative in nature and 
consequently does not affect the safety analysis in any way. 
Consequently, the margin of safety is not being reduced by the 
proposed Technical Specifications change.

    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: 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.

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

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

Carolina Power & Light Company, et al., Docket No. 50-325, 
Brunswick Steam Electric Plant, Unit 1, Brunswick County, North 
Carolina

    Date of amendment request: April 14, 2000.
    Description of amendment request: The proposed amendment would 
modify Surveillance Requirement 3.1.3.3 to allow partial insertion of 
control rod 26-47 instead of insertion of one complete notch. This 
revised acceptance criterion would be limited to the current Unit No. 1 
operating cycle, after which the current one-notch requirement will be 
re-established.
    Date of publication of individual notice in Federal Register: April 
21, 2000 (65 FR 21481).
    Expiration date of individual notice: May 22, 2000.

GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear 
Station, Unit 2, Middletown, Pennsylvania

    Date of amendment request: April 6, 2000.
    Brief description of amendment request: The proposed amendment 
would reflect an administrative name change from GPU Nuclear 
Corporation to GPU Nuclear, Inc. Furthermore, the proposed license 
amendment makes an editorial change to better describe TMI-2's use of 
site physical security, guard training and qualification, and safeguard 
contingency plans that are maintained by the Three Mile Island Nuclear 
Station, Unit 1, licensee, AmerGen Energy Company, LLC. In addition, 
the licensee requests that minor changes (mainly in titles) be made in 
Section 6.0 of the Technical Specifications to reflect the TMI-2 
organizational and administrative controls that will exist following 
the sale of the Oyster Creek Nuclear Generating Station.
    Date of publication of individual notice in Federal Register: April 
21, 2000 (65 FR 21484).
    Expiration date of individual notice: May 21, 2000.

Notice of Issance 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

[[Page 25771]]

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).

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert 
Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

    Date of application for amendment: February 18, 2000, as 
supplemented March 3, 2000.
    Brief description of amendment: The amendment approved resolution 
of an issue involving the Societie Alsacienne Construction Mechaniques 
Del Melhouse (SACM) diesel generator (DG) that constitutes an 
unreviewed safety question. Specifically, a new failure mode has been 
identified for DG 1A SACM that is not adequately described in the 
Updated Final Safety Analysis Report. The manufacturer has indicated 
that operating the engine in a light load condition may degrade engine 
performance and ultimately result in engine failure.
    Date of issuance: April 20, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 235.
    Renewed Facility Operating License No. DPR-53: Amendment revised 
the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: March 7, 2000 (65 FR 
12038).
    The March 3, 2000, submittal 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 20, 2000.
    No significant hazards consideration comments received: No.

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: May 27, 1999, as supplemented by 
letters of February 25, 2000, March 30, 2000, and eMail of March 13, 
2000.
    Brief description of amendment: This amendment revises the maximum 
allowable service water temperature permitted by Surveillance 
Requirement 3.7.8.2 for the ultimate heat sink (UHS) from the currently 
permitted limit of 95  deg.F to 97  deg.F while it restores the 
original Technical Specifications provisions for required action and 
completion times of 6/36 hours to be in mode 3/5, respectively, in the 
event the UHS temperature were to exceed 97  deg.F.
    Date of issuance: April 18, 2000.
    Effective date: April 18, 2000.
    Amendment No.: 187.
    Facility Operating License No. DPR-23: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 23, 2000 (65 
FR 9001). The supplements of February 25, March 13, and March 30, 2000, 
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 18, 2000.
    No significant hazards consideration comments received: No

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: May 5, 1999, as supplemented on 
October 8, 1999.
    Brief description of amendments: The amendments resolved an 
Unreviewed Safety Question (USQ) related to an evaluation of the 
reactor building ventilation system exhaust plenum masonry walls. The 
amendments approved the use of different methodology and acceptance 
criteria for the reassessment of certain masonry walls subjected to 
transient pressurization loads resulting from a high energy line break. 
This change to the licensing basis, when evaluated by the licensee in 
accordance with 10 CFR 50.59, resulted in an USQ that required prior 
approval by the NRC staff in accordance with the provisions of 10 CFR 
50.90.
    Date of issuance: April 11, 2000.
    Effective date: Immediately, to be implemented during the next 
scheduled Final Safety Analysis Report update.
    Amendment Nos.: 139 and 124.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: June 16, 1999 (64 FR 
32286). The October 8, 1999, submittal provided additional clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 11, 2000.
    No significant hazards consideration comments received: No.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: September 23, 1999.
    Brief description of amendment: The amendment relocates items 
associated with instrumentation for toxic gas monitoring from Technical 
Specifications to the Updated Final Safety Analysis Report.
    Date of issuance: April 20, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 208
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 1, 1999 (64 FR 
67332).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 20, 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 22, 1999, as supplemented by 
letters dated March 20, March 24 (2), March 29, and April 5, 2000.
    Brief description of amendment: The amendment authorized revisions 
to the radiological assessment calculational methodology for the loss-
of-coolant accident (LOCA) and the control rod drop accident (CRDA). 
The amendment request was submitted to address potential unreviewed 
safety questions resulting from these revisions due to instances of 
increased dose consequences. Because of outstanding issues involving 
various assumptions used in these calculational methodologies, the 
staff is deferring the review of implementing this change on a 
permanent basis. Subsequently, this amendment is to be effective 
immediately and remain effective until Cooper Nuclear Station enters 
mode 4 in preparation for refueling outage 20 (effectively, one 
operating cycle). Also, the staff has deferred review of the 
radiological assessment methodology revisions for the fuel handling 
accident (FHA) and the main steamline break (MSLB) accident. It is 
anticipated that Nebraska Public Power District (NPPD) will resolve any 
outstanding issues concerning these calculational methodology revisions 
in a timely manner in support of a permanent change that is acceptable 
to the staff.

[[Page 25772]]

    Date of issuance: April 7, 2000.
    Effective date: April 7, 2000, to be implemented within 30 days and 
remain effective until Cooper Nuclear Station enters mode 4 in 
preparation for refueling outage 20.
    Amendment No.: 183.
    Facility Operating License No. DPR-46: The amendment authorizes 
changes to the licensing basis and changes to the operating license.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4280). The March 20 and 24 (2), March 29, and April 5, 2000, letters 
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 April 7, 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 15, 1999, as supplemented by 
letters dated February 15 and April 8, 2000.
    Brief description of amendment: The technical specification change 
revises the average power range monitors (APRMs) neutron flux-high 
(flow-biased) allowable value based on a revised power-to-flow map. The 
revised power-to-flow map extends the current plant operating domain to 
above the rated rod line, to within an envelope referred to as the 
maximum extended load line limit (MELLL) and adds the increased core 
flow (105 percent) region.
    Date of issuance: April 11, 2000.
    Effective date: April 11, 2000, to be implemented within 30 days.
    Amendment No.: 184.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4279). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 11, 2000.
    No significant hazards consideration comments received: No.

North Atlantic Energy Service Corporation, et al., Docket No. 50-
443, Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: November 19, 1999.
    Description of amendment request: The amendment revised Technical 
Specification (TS) 6.4.3, ``Nuclear Safety Audit Review Committee 
(NSARC),'' by relocating the specific requirements of this TS to the 
Quality Assurance Program located in the Updated Final Safety Analysis 
Report (UFSAR).
    Date of issuance: April 11, 2000.
    Effective date: As of its date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 67.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications and authorized changes to the UFSAR.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4281).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 11, 2000.
    No significant hazards consideration comments received: No.

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 1, 2000.
    Brief description of amendment: The amendment revises limiting 
conditions for operation (LCO) 3.0.1 and 3.0.2 and adds LCO 3.0.5 to 
the Technical Specifications (TSs) for Milstone 3. LCO 3.0.5 
establishes allowances for restoring equipment to service under 
administrative controls when the equipment has been removed from 
service or declared inoperable to comply with actions in the TSs.
    Date of issuance: April 17, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 179.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 1, 2000 (65 FR 
11092).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 17, 2000.
    No significant hazards consideration comments received: No.

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: November 10, 1999, as 
supplemented February 25, 2000.
    Brief description of amendments: The amendments revise the elevated 
F-star (EF*) distance for the steam generator tubes specified in 
Technical Specification 4.12.D.1.(l) following a correction to a minor 
error in the calculations supporting the current EF* distance.
    Date of issuance: April 19, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of the date of issuance.
    Amendment Nos.: 149 and 140.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 23, 2000 (65 
FR 9010).
    The February 25, 2000, supplemental letter provided clarifying 
information that did not change the staff's initial proposed no 
significant hazards consideration determination and did not expand it 
beyond the scope of the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 19, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 7, 1999, as supplemented 
September 27, 1999.
    Brief description of amendments: Revised Technical Specifications 
Section 3/4.4.3 to clarify the action statement concerning inoperative 
reactor coolant leakage detection systems.
    Date of issuance: April 5, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 140 and 103.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38034).
    The September 27, 1999, letter provided clarifying information that 
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 5, 2000.
    No significant hazards consideration comments received: No.

PECO Energy Company, Docket No. 50-352, Limerick Generating 
Station, Unit 1, Montgomery County, Pennsylvania

    Date of application for amendment: October 14, 1999, as 
supplemented February 11, 2000.
    Brief description of amendment: This amendment revised Technical 
Specification (TS) Section 2.2, ``Safety

[[Page 25773]]

Limits and Limiting Safety Systems Settings,'' and TS Section 3.0/4.0, 
``Limiting Conditions for Operation and Surveillance Requirements.''
    Date of issuance: April 12, 2000.
    Effective date: Effective as of the date of issuance and; Unit 1 
shall be implemented during the LGS Unit 1 refueling outage scheduled 
to begin March 29, 2000.
    Amendment No.: 141.
    Facility Operating License No. NPF-39. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 1, 1999 (64 FR 
67337).
    The February 11, 2000, letter 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 12, 2000.
    No significant hazards consideration comments received: No.

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: February 26, 1998, as 
supplemented October 14, 1999.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TSs) by changing the value of the allowable containment 
leakage rate to 1.5 percent per day and correcting conflicting 
information in TS Section 4.6.C, ``Coolant Chemistry.''
    Date of issuance: April 14, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 261.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19977).
    The October 14, 1999, 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 14, 2000.
    No significant hazards consideration comments received: No.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. 
Ginna Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: November 30, 1999.
    Brief description of amendment: The amendment revises Technical 
Specification 5.5.10, ``Ventilation Filter Testing Program'' to meet 
the actions requested by Generic Letter 99-02.
    Date of issuance: April 12, 2000.
    Effective date: April 12, 2000.
    Amendment No.: 77.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4290)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 12, 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: February 26, 1999.
    Brief description of amendments: These amendments revised the 
Technical Specifications (TSs) to eliminate inconsistencies and 
redundancies in Section 3.8.1.1, action statements involving inoperable 
offsite AC circuits and combinations of inoperable offsite power 
supplies and emergency diesel generators.
    Date of issuance: April 14, 2000.
    Effective date: April 14, 2000.
    Amendment Nos.: 255 and 246.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: March 24, 1999 (64 FR 
14287).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 14, 2000.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of no Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have

[[Page 25774]]

been issued and made effective 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 
application for amendment, (2) the amendment to Facility Operating 
License, 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).
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By June 2, 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. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    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 the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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: The amendments revised the 
Technical Specifications (TS) requirements applicable to opening of 
secondary containment access doors.
    Date of issuance: April 21, 2000.
    Effective date: April 21, 2000.
    Amendment Nos.: 238, 264, and 224.
    Facility Operating License No. DPR-33, DPR-52 and DPR-68: 
Amendments revise the TS.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes (65 FR 18141 dated April 6, 2000). The notice 
provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. No comments have been received. The notice also 
provided for an opportunity to request a hearing by April 20, 2000, but 
indicated that if the Commission makes a final NSHC determination, any 
such hearing would take place after issuance of the amendment.

[[Page 25775]]

    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and final determination of NSHC are contained in 
a Safety Evaluation dated April 21, 2000.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

    Dated at Rockville, Maryland, this 26th 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. 00-10743 Filed 5-2-00; 8:45 am]
BILLING CODE 7590-01-P