[Federal Register Volume 64, Number 154 (Wednesday, August 11, 1999)]
[Notices]
[Pages 43764-43785]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-20545]


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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 July 17, 1999, through July 30, 1999. The 
last biweekly notice was published on July 28, 1999 (64 FR 40903).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-

[[Page 43765]]

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 September 10, 1999, the licensee may file a request for a 
hearing with respect to issuance of the amendment to the subject 
facility operating license and any person whose interest may be 
affected by this proceeding and who wishes to participate as a party in 
the proceeding must file a written request for a hearing and a petition 
for leave to intervene. Requests for a hearing and a petition for leave 
to intervene shall be filed in accordance with the Commission's ``Rules 
of Practice for Domestic Licensing Proceedings'' in 10 CFR part 2. 
Interested persons should consult a current copy of 10 CFR 2.714 which 
is available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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: July 9, 1999.
    Description of amendment request: The proposed amendment would 
revise Harris Nuclear Plant (HNP) Technical Specification (TS) 3/4.2.2, 
``Heat Flux Hot Channel Factor--FQ(Z),'' TS 3/4.2.3, ``RCS 
Flow Rate And Nuclear Enthalpy Rise Hot Channel Factor,'' TS 3/4.2.5, 
``DNB Parameters,'' an associated note in TS Table 2.2-1, and 
associated Bases. Specifically, the proposed amendment would: (1) 
Remove the allowance for reduced power operation for reduced Reactor 
Coolant System (RCS) flow rate conditions; (2) separate the 
requirements for F delta H and RCS flow rate in the format prescribed 
by NUREG-1431, Revision 1, ``Standard Technical Specifications, 
Westinghouse Plants,'' dated April 1995; and, (3) implement the 
guidance of NUREG-1431, Revision 1, and NRC Generic Letter (GL) 88-16, 
dated October 4, 1988 for TS 3/4.2.2, TS 3/4.2.3, TS 3/4.2.5 and 
associated Bases by removing cycle specific parameters and placing that

[[Page 43766]]

information into the Core Operating Limits Report (COLR).
    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 amendment will not introduce any new equipment or 
require existing equipment to function different from that 
previously evaluated in the Final Safety Analysis Report (FSAR) or 
TS.
    As described in HNP TS Bases, the limits on heat flux hot 
channel factor, RCS flow rate, and enthalpy rise hot channel factor 
ensure that: (1) the design limits on peak local power density and 
minimum DNBR [departure from nucleate boiling ratio] are not 
exceeded and (2) in the event of a LOCA the peak fuel clad 
temperature will not exceed the 2200 degree Fahrenheit ECCS 
[emergency core cooling system] acceptance limit.
    Removing the allowance for reduced power operation for reduced 
RCS flow conditions is more restrictive than that currently allowed 
by TS. Power Distribution Limiting Conditions for Operation for heat 
flux hot channel factor and enthalpy rise hot channel factor are not 
affected by this change. Therefore, the consequences of an accident 
will not increase because of this change. Power Distribution limits 
place administrative restrictions on reactor core parameters and as 
such do not initiate nor mitigate accidents.
    Power Distribution limits at HNP are developed using NRC 
approved methodologies. Changing power distribution limits to be 
consistent with NUREG-1431, Revision 1 will not increase the 
probability or consequences of an accident that has been previously 
evaluated.
    Relocating cycle specific information from TS to the COLR will 
not impact the ability of structures, systems, or components to 
mitigate accidents. Future changes to relocated requirements in the 
COLR will be submitted to the NRC for review in accordance with HNP 
TS Section 6.
    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.
    The proposed amendment will not introduce any new equipment or 
require existing equipment to function different from that 
previously evaluated in the Final Safety Analysis Report (FSAR) or 
TS. The changes are consistent with NUREG-1431, Revision 1 and the 
Commission's Final Policy Statement on Technical Specification 
improvements. The proposed amendment will not create any new 
accident scenarios, because the change does not introduce any new 
single failures, adverse equipment or material interactions, or 
release paths.
    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 LCO limit for RCS flow rate at 100.0% reactor power has not 
changed. The previous capability to operate with reduced RCS flow 
rate has been eliminated. This aspect of the proposed change is more 
restrictive than current plant TS in that continued reactor 
operation greater than 5% is not allowed if RCS flow rate is less 
than the LCO limit at 100% power.
    Changes to TS 3/4.2.2, TS 3/4.2.3, TS 
3/4.2.5 and associated Bases are in accordance with NUREG-1431, 
Revision 1. The completion times for TS Actions are acceptable 
because the plant is not allowed to remain in an unacceptable 
condition for an extended period of time. Sufficient time to reduce 
reactor power in an orderly manner or perform other required actions 
is also provided. The surveillance intervals established by NUREG-
1431, Revision 1 have been determined to be adequate for monitoring 
the change in power distribution.
    Relocating cycle specific information from HNP TS to the COLR is 
in accordance with NRC GL 88-16. HNP does not intend to alter the 
methodologies for any parameter limit calculation as a result of 
this change. The proposed change is in accordance with the plant 
safety analysis. 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.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Section Chief: Sheri R. Peterson.

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: July 9, 1999.
    Description of amendment request: The proposed amendment would 
relocate Harris Nuclear Plant (HNP) Technical Specification (TS) 3/
4.3.3.3, ``Seismic Instrumentation,'' TS 
3/4.3.3.4, ``Meteorological Instrumentation,'' TS 3/4.3.3.9, ``Metal 
Impact Monitoring System,'' and TS 
3/4.3.3.11, ``Explosive Gas Monitoring Instrumentation,'' to plant 
procedure PLP-114, ``Relocated Technical Specifications and Design 
Basis Requirements.'' The proposed change is in accordance with 
guidance provided by NRC Generic Letter 95-10, ``Relocation of Selected 
Technical Specification Requirements Related to Instrumentation.'' 
Changes to relocated requirements would be performed in accordance with 
10 CFR 50.59. Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Seismic Instrumentation, Meteorological Instrumentation, Metal 
Impact Monitoring System, and Explosive Gas Monitoring 
Instrumentation are not accident initiating components as described 
in the Final Safety Analysis Report. Seismic Instrumentation, 
Meteorological Instrumentation, Metal Impact Monitoring System, and 
Explosive Gas Monitoring Instrumentation are not accident mitigating 
components. There are no modifications being made to plant systems 
as a result of this change. Additionally, there are no changes being 
made to the way in which systems are being operated as a result of 
this change. 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.
    Seismic Instrumentation, Meteorological Instrumentation, Metal 
Impact Monitoring System, and Explosive Gas Monitoring 
Instrumentation are not accident initiating components as described 
in the Final Safety Analysis Report (FSAR). The proposed change 
relocates the TS requirements for Seismic Instrumentation, 
Meteorological Instrumentation, Metal Impact Monitoring System, and 
Explosive Gas Monitoring Instrumentation to plant procedure PLP-114. 
Plant systems and components are not modified as a result of this 
change. Future changes in these systems will be controlled in 
accordance with 10 CFR 50.59.
    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 Seismic Instrumentation, Meteorological 
Instrumentation, Metal Impact Monitoring System, and Explosive Gas 
Monitoring Instrumentation does not affect any of the

[[Page 43767]]

parameters that relate to the margin of safety as described in the 
Bases of the TS or the FSAR. Accordingly, NRC Acceptance Limits are 
not affected by this change. The proposed change relocates the TS 
requirements for Seismic Instrumentation, Meteorological 
Instrumentation, Metal Impact Monitoring System, and Explosive Gas 
Monitoring Instrumentation to plant procedure PLP-114. Plant systems 
and components are not modified as a result of this change. Future 
changes in these systems will be controlled in accordance with 10 
CFR 50.59. Generic Letter 95-10 states that the staff has concluded 
that these provisions are not related to dominant contributors to 
plant risk.
    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.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Section Chief: Sheri R. Peterson.

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

    Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 
1 and 2, Will County, Illinois.
    Date of amendment request: June 30, 1999
    Description of amendment request: The proposed amendment would 
clarify that the source of DC electrical power required for a unit in 
Mode 5 or 6 or during the movement of irradiated fuel assemblies may be 
cross-tied to the opposite unit. An administrative change would also 
delete reference to AT&T batteries since all AT&T batteries have been 
replaced with Charter Power Systems, Inc. (C&D) batteries. The 
amendment would also remove the Allowed Outage Time (AOT) extension 
approved for Braidwood Station by Amendment No. 99. The activity 
addressed by Amendment No. 99 is complete and the extension no longer 
applies.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change will allow one DC bus on a shutdown unit to 
be supplied via the DC bus cross-tie to the opposite unit. The other 
DC bus on the shutdown unit will at all times be required to be 
fully operable, supplied by the associated battery and charger, and 
the associated cross-ties open. The DC electrical system is not 
considered an initiator of any accident previously evaluated, and 
therefore the probability of a previously analyzed accident is 
unchanged.
    The consequences of a previously analyzed event are dependent on 
the initial conditions assumed for the analysis, the availability 
and successful functioning of the equipment assumed to operated in 
response to the analyzed event, and the setpoints at which these 
actions are initiated. Sufficient equipment remains available to 
mitigate the consequences of previously analyzed events. The Updated 
Final Safety Analysis Report (UFSAR) section 8.3.2.1.1 clearly 
allows operation with the DC cross-tie closed on one DC bus between 
a unit that is operating and a unit that is shutdown, or between two 
shutdown units, in the manner proposed by this amendment. The TS in 
effect prior to the implementation of the Improved TS also allowed 
operation in the manner proposed by this amendment. If DC buses are 
cross-tied due to an inoperable DC source on a shutdown unit, both 
the previous TS and the change proposed by this amendment limit the 
time in this condition to seven days, and if the inoperable source 
is a battery, the current on the cross-tie is limited to 200 amps. 
These actions protect both the operating unit, and the shutdown 
unit. If a shutdown unit's DC bus is cross-tied to an operating 
unit's DC bus due to an inoperable charger on the operating unit, 
both the previous TS and the change proposed by this amendment limit 
the time in this condition to 24 hours. The limitations imposed by 
both the previous TS and the change proposed by this amendment 
ensure that operation in this configuration is within the design 
bases of the plant. Thus the consequences of accidents previously 
analyzed are unchanged between the previous TS and the change 
proposed by this amendment. In the worst case scenario, assuming a 
single failure, one DC bus on the shutdown unit will always be 
operable, and the ability to mitigate the consequences of any 
accident previously analyzed is preserved.
    The change to delete all references in the Braidwood TS to AT&T 
batteries and the AOT extension granted under TS Amendment Number 99 
is administrative only, and has no impact on the probability or 
consequences of accidents previously evaluated.
    Therefore this proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed 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 change to the 
plant. No new equipment is being introduced, and installed equipment 
is not being operated in a new or different manner. There is no 
change being made to the parameters within which the plant is 
operated. There are no setpoints affected by this change at which 
protective or mitigative actions are initiated. This 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 which ensure the plant remains within analyzed 
limits in being proposed, and no change is being made to the 
procedures relied upon to respond to an off-normal event. As such, 
no new failure modes are being introduced. The change does not alter 
assumptions made in the safety analysis and licensing basis. 
Therefore, the change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The change to delete all references in the Braidwood TS to AT&T 
batteries and the AOT extension granted under TS Amendment Number 99 
is administrative only, and cannot create the possibility of a new 
or different kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. Sufficient equipment remains available to actuate 
upon demand for the purpose of mitigating an analyzed event. The 
proposed change, which will allow one DC bus on a shutdown unit to 
be supplied via the DC bus cross-tie to the opposite unit, is 
acceptable because of the limitations imposed on operation in this 
configuration, and because the other DC bus on the shutdown unit 
will at all times be required to be fully operable, supplied by the 
associated battery and charger, and the associated cross-ties open. 
The TS in effect prior to the implementation of the Improved TS 
allowed operation in the manner proposed by this amendment. In the 
worst case scenario, assuming a single failure, one DC bus on the 
shutdown unit will always be operable. Thus, there is no detrimental 
impact on any equipment design parameter, and the plant will still 
be required to operate within prescribed limits. Therefore, the 
change does not reduce the margin of safety.
    The change to delete all references in the Braidwood TS to AT&T 
batteries and the AOT extension granted under TS Amendment Number 99 
is administrative only, and does not reduce the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff

[[Page 43768]]

proposes to determine that the requested amendments involve no 
significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: May 3, 1999.
    Description of amendment request: The proposed amendments would 
relocate Technical Specifications (TS) Section 3/4.6.I to the Updated 
Final Safety Analysis Report (UFSAR). TS Section 3/4.6.I contains 
reactor coolant chemistry limiting conditions for operation (LCO) and 
surveillance requirements (SR) for conductivity, chloride 
concentration, and pH.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes simplify the TS, meet regulatory 
requirements for relocated TS's, and implement the recommendations 
of the NRC Final Policy Statement on TS improvements. The Chemistry 
requirements will be relocated to the Updated Final Safety Analysis 
Report (UFSAR) and to applicable station procedures. Future changes 
to these requirements will be controlled by 10 CFR 50.59. The 
proposed changes are administrative in nature and do not involve any 
modification to any plant equipment or affect plant operation. 
Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any previously 
evaluated accident.
    Consequently, this proposed amendment does not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes are administrative in nature, do not 
involve any physical alterations to any plant equipment, and cause 
no change in the method by which any safety related system performs 
its function. Therefore, this proposed TS amendment will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed amendment represents the relocation of current 
requirements, which are based on generic guidance or previously 
approved provisions for other stations. The proposed changes are 
administrative in nature and do not adversely affect existing plant 
safety margins or the reliability of the equipment assumed to 
operate in the safety analysis. The proposed changes have been 
evaluated and found to be acceptable for use at Dresden Nuclear 
Power Station. Since the proposed changes are administrative in 
nature, and are based on NRC accepted provisions which have been 
adopted at other nuclear facilities, and maintain the necessary 
levels of system reliability, the proposed changes do not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

Commonwealth Edison Company, Docket No. 50-373, LaSalle County Station, 
Unit 1, LaSalle County, Illinois

    Date of amendment request: July 7, 1999.
    Description of amendment request: The proposed amendments would (1) 
revise Technical Specification Section 2.1, Safety Limits, to reflect a 
change to the LaSalle, Unit 1, Minimum Critical Power Ratio Safety 
Limit; and (2) revise Technical Specification Section 6.6.A.6 to add an 
NRC-approved Siemens Power Corporation methodology to the list of 
topical reports used to determine the core operating limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. Limits have been established consistent with NRC-
approved methods to ensure that fuel performance during normal, 
transient, and accident conditions is acceptable. These changes do 
not affect the operability of plant systems, nor do they compromise 
any fuel performance limits.
    Changing the MCPR Safety Limit for LaSalle Unit 1 will not 
increase the probability or the consequences of an accident 
previously evaluated. This change implements the MCPR Safety Limit 
resulting from the SPC ANFB critical power correlation methodology 
using the approved ATRIUM-9B additive constant uncertainty. For each 
cycle, cycle specific MCPR Safety Limit calculations will be 
performed, consistent with SPC's approved methodology, to confirm 
the appropriateness of the MCPR Safety Limit. Additionally, 
operational MCPR limits will be applied that will ensure the MCPR 
Safety Limit is not violated during all modes of operation and 
anticipated operational occurrences. The MCPR Safety Limit ensures 
that less than 0.1% of the rods in the core are expected to 
experience boiling transition. Therefore the probability or 
consequences of an accident will not increase.
    Adding EMF-85-74, Revision 0, Supplement 1 (P)(A) and Supplement 
2 (P)(A) to Section 6 does not increase the probability or 
consequences of an accident previously evaluated. The NRC-approved 
burnup extension for RODEX2A applications has been demonstrated to 
meet all applicable design criteria. Therefore adding this 
methodology to Technical Specification Section 6 does not increase 
the probability or consequences of an accident previously evaluated 
.
    Therefore, this proposed amendment 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?
    Creation of the possibility of a new or different kind of 
accident would require the creation of one or more new precursors of 
that accident. New accident precursors may be created by 
modifications to the plant configuration, including changes in 
allowable modes of operation. This Technical Specification submittal 
does not involve any modifications to the plant configuration or 
allowable modes of operation. No new precursors of an accident are 
created and no new or different kinds of accidents are created. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Changing the MCPR Safety Limit does not create the possibility 
of a new accident from any accident previously evaluated. This

[[Page 43769]]

change does not alter or add any new equipment or change modes of 
operation. The MCPR Safety Limit is established to ensure that 99.9% 
of the rods avoid boiling transition.
    The MCPR Safety Limit is changing for LaSalle Unit 1 to support 
Cycle 9 operation. This change does not introduce any physical 
changes to the plant, alter the processes used to operate the plant, 
or change allowable modes of operation. Therefore, no new accidents 
are created that are different from any accident previously 
evaluated.
    The addition of RODEX2A (EMF-85-74, Revision 0, Supplement 1 
(P)(A) and Supplement 2 (P)(A)) does not create the possibility of a 
new accident from an accident previously evaluated. This change does 
not alter or add any new equipment or change modes of operation. 
This change does not introduce any physical changes to the plant, 
alter the processes used to operate the plant, or change allowable 
modes of operation. Therefore, no new accidents are created that are 
different from any accident previously evaluated.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the change involve a significant reduction in the margin 
of safety?
    Changing the MCPR Safety Limit for LaSalle Unit 1 will not 
involve any reduction in margin of safety. The MCPR Safety Limit 
provides a margin of safety by ensuring that less than 0.1% of the 
rods are calculated to be in boiling transition. The proposed 
Technical Specification amendment request reflects the MCPR Safety 
Limit results from evaluations by SPC using NRC-approved 
methodology.
    The revised MCPR Safety Limit will ensure the same level of fuel 
protection. Additionally, operational limits will be established 
based on the proposed MCPR Safety Limit to ensure that the MCPR 
Safety Limit is not violated during all modes of operation including 
anticipated operation[al] occurrences. This will ensure that the 
fuel design safety criterion of more than 99.9% of the fuel rods 
avoiding transition boiling during normal operation as well as 
during an anticipated operational occurrence is met.
    The addition of EMF-85-74, Revision 0, Supplement 1 (P)(A) and 
Supplement 2 (P)(A) to Section 6 does not decrease the margin of 
safety. The burnup limit extension for RODEX2A applications has been 
reviewed and approved by the NRC. The data supporting the burnup 
extension demonstrates that all applicable design criteria are met. 
Therefore, since the burnup extension is acceptable and within the 
design criteria, using the approved burnup extension will not affect 
the margin of safety.
    Therefore, these changes do not involve a significant reduction 
in the margin of safety.
    Therefore, based upon the above evaluation, ComEd has concluded 
that these changes involve no significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 815 
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby, 
Illinois 61348-9692.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: November 9, 1998, as supplemented on 
July 7, 1999.
    Description of amendment request: The proposed amendments would 
revise Technical Specification Table 3.3.3-2, ``Emergency Core Cooling 
System Actuation Instrumentation Setpoints,'' to modify the degraded 
voltage second level undervoltage relay setpoint and allowable value. 
These proposed amendments were originally noticed on January 13, 1999 
(64 FR 2245), and are being renoticed to include the revised setpoints 
that were included in the July 7, 1999, supplement.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The setpoint change does not change the logic or function of the 
degraded voltage protection circuits as described in the UFSAR 
[Updated Final Safety Analysis Report] Section 8.2.3. They also do 
not reduce the reliability of these circuits. The increase in the 
degraded voltage protection circuit setpoint is conservative 
compared to the existing setpoint. There is no change as a result of 
this amendment to the underlying accident and transient analyses 
that support operations of LaSalle County Station. Inadvertent or 
spurious operation of the degraded voltage protection function will 
initiate loading of the safe shutdown loads on the diesel generators 
and is not assumed to initiate an accident. The proposed degraded 
voltage setpoints are low enough to prevent spurious actuations 
given the expected offsite grid voltages. After implementation of 
this amendment, no operator actions are required for equipment 
operations in response to degraded voltage conditions.
    This change does not affect the initiators or precursors of any 
accident previously evaluated. This change will not increase the 
likelihood that a transient initiating event will occur because 
transients are initiated by equipment malfunction and/or 
catastrophic system failure.
    The consequences of accidents previously evaluated are not 
increased. The proposed change does not affect the required level of 
availability of systems required to mitigate the accidents 
considered in the analyses. The proposed changes will ensure that 
the Class 1E equipment will be capable of starting and operating 
during a design basis accident with degraded offsite grid voltage. 
The increase in the level of confidence is the result of more 
rigorous methodology used to determine limiting Class 1E bus 
voltages at the minimum expected offsite AC voltage. These 
calculations demonstrate that the degraded voltage relays will not 
actuate following a block start of the electrical loads that are 
automatically actuated by or as a consequence of the LOCA [loss-of-
coolant accident] signal if the switchyard voltage remains above 352 
kV.
    If the grid voltage drops below 352 kV, then the analytical 
limit of 3814 volts for proper operation of class 1E loads connected 
to each 4.16 kV Class 1E bus is assured by transfer to the 
respective onsite power sources (Emergency Diesel Generators (EDGs)) 
by the degraded voltage logic.
    Therefore this proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated because:
    Setpoint methodology established the bases to ensure that, with 
known errors, the relays will detect degraded voltage conditions and 
transfer safety loads to the EDGs at a voltage level adequate to 
ensure proper safety equipment performance and to prevent equipment 
damage.
    The trip setpoint of greater than or equal to 3863 volts and 
less than or equal to 3877 volts and the allowable value of greater 
than or equal to 3814 volts and less than or equal to 3900 volts, 
include adequate tolerance to calibrate the relay trip units while 
ensuring that the Class 1E bus voltage will remain above the 
analytical limits.
    These setpoint changes will ensure that adequate voltages will 
be available for the continuous operation of safety-related 
equipment required to function during a LOCA. These proposed changes 
will also ensure that adequate voltages will be available for 
starting any Class 1E equipment.
    The proposed degraded voltage setpoint change does not change 
the design of the degraded voltage protection system or its function 
to protect against degraded offsite power. Actuation of the degraded 
voltage protection system will initiate a sequence of events that 
will start the EDG for the associated Class 1E bus, strip loads from 
the Class 1E bus, open all feed breakers to the Class 1E bus, close 
the Emergency feed breaker (thus energizing the Class 1E bus

[[Page 43770]]

from the respective EDG), and initiate starting of the Safe Shutdown 
equipment supplied by the Class 1E bus.
    Since the scope of this change does not affect the operation of 
auxiliary power system or any actions necessary to mitigate the 
consequences of accidents or achieve safe shutdown, the change does 
not involve a new or different accident scenario.
    Therefore, these proposed changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    (3) Involve a significant reduction in the margin of safety 
because:
    The proposed amendment will allow the degraded voltage setpoint 
to be conservatively established based on new engineering 
calculations which consider the lowest expected offsite grid voltage 
and operation of required Class 1E equipment under design basis 
accident loading conditions.
    The proposed degraded voltage setpoints will ensure that 
adequate Class 1E bus voltage will be available to support starting 
and operation of required Class 1E loads. The proposed setpoint 
includes instrument error to ensure that the lowest possible voltage 
will not be lower than the degraded voltage analytical limits. 
Additionally, the proposed setpoints are low enough to prevent 
spurious actuations due to expected fluctuations in the grid 
voltage. The new setpoints are also set with margin to the minimum 
Class 1E bus voltage, which is based on a minimum grid voltage of 
352 kV, which is less than the expected grid voltage of 354 kV. The 
proposed changes will provide an increase in the level of protection 
that currently exists and will ensure the margin of safety is 
adequately maintained.
    Therefore, these changes do not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 815 
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby, 
Illinois 61348-9692.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: April 20, 1999 (Reference NRC-99-0035).
    Description of amendment request: The proposed amendment will 
revise the Technical Specifications by deleting Specification D.3.c. 
Specification D.3.c requires the licensee to perform weekly 
observations of the nitrogen cover gas pressure within the sodium 
storage tanks located in the Sodium Building Complex. Removing this 
surveillance requirement would allow the licensee to remove the 
nitrogen cover gas system from service for these sodium storage tanks. 
This action is necessary for the licensee to begin work on removing the 
remaining residual sodium from these tanks. The licensee also requested 
an editorial change to delete the words ``STORAGE TANK'' from the title 
of Specification D.
    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 previously 
evaluated.
    Removing the primary cover gas supply from the storage tanks 
will not significantly increase the probability of an accident 
occurring as long as the probability of an uncontrolled water 
reaction with residual sodium is not significantly increased. This 
is ensured by sealing the storage tanks after the nitrogen cover gas 
system is removed except when controlled activities such as sampling 
are performed. The consequences of an accident would not be affected 
by removing the nitrogen cover gas supply from service as the 
previously analyzed primary sodium accident already involves release 
of all the radioactive material in the primary sodium. Removing the 
cover gas 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.
    A sodium accident has been previously evaluated. No other type 
of accident could be caused by removing the primary sodium tanks 
cover gas or opening the tanks since no other system or mode of 
operation of any other system will be affected.
    (3) The proposed change does not involve a significant reduction 
in a margin of safety.
    Currently, only a small amount of residual sodium remains in the 
primary sodium storage tanks. Some of this residual sodium may have 
been converted to sodium carbonate. This conversion of sodium to 
sodium carbonate would have left even less sodium remaining in these 
tanks. The cover gas is a good precaution, especially for tanks 
sitting unattended for many years. It prevents moisture from 
intruding into the tanks and reacting with the sodium residues. It 
also prevents oxygen from entering these tanks 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 tank in an uncontrolled manner. Even if slight 
amounts of moisture from humidity in the air enter these tanks over 
the next year or two, until the sodium is removed while the tanks 
are either opened or sealed, the volume of each tank (15,000 
gallons) is large enough that the tank should be able to dissipate 
any small reactions that could occur. The design pressure for the 
primary sodium storage tanks is from vacuum to 50 pounds per square 
inch based on the vendor's drawing.
    Even if sufficient water entered the tank, generated hydrogen, 
and sufficient oxygen entered the tank to cause a reaction that 
released the contents of the tank, there would be no significant 
release of radioactivity from the tank. The release of all residual 
primary sodium would result in concentration levels well below the 
values in 10 CFR 20, Appendix B, Table II for releases to 
unrestricted areas. Since there is less sodium in the primary sodium 
storage tanks than in the secondary sodium storage tanks, potential 
hazard consequences of releasing the contents of a primary sodium 
tank are bounded by the hypothetical secondary sodium scenario 
evaluated in the Fermi 1 Safety Analysis Report. For these reasons, 
the proposed change does not involve a significant reduction in the 
margin of safety.

    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.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.
    Attorney for licensee: John Flynn, Esquire, Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Branch Chief: Larry W. Camper.

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

    Date of amendment request: July 22, 1998, supplemented by October 
22, 1998, January 28, May 6 and June 24, 1999.
    Description of amendment request: By the referenced submittals the 
licensee requested the Catawba Technical Specifications be changed to 
permit the licensee's planned use of fuel supplied by Westinghouse, 
which has different design characteristics from the fuel currently in 
use. The staff has previously published two Notices of Consideration of 
Issuance of Amendments and Proposed No Significant Hazards 
Consideration of Issuance of Amendments. The first notice, dated 
November 18, 1998 (63 FR 64108), covers the submittals dated July

[[Page 43771]]

22 and October 22, 1998. The second notice, dated May 19, 1999 (64 FR 
27317), covers the submittal dated May 6, 1999. The June 24, 1999, 
submittal actually requested an amendment separate from that described 
above, but nevertheless conveyed a revised proposed Figure 2.1.1-1, 
``Reactor Core Safety Limits--Four Loops in Operation'', superseding 
what was originally proposed in the licensee's previous submittals. 
Hence, this Notice only covers the revised proposed Figure 2.1.1-1. The 
Notices referenced above are unaffected.
    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 for the June 24, 1999, submittal. The staff has reviewed 
the licensee's analysis and has performed its own analysis as follows:

First Standard

    No. The proposed changes to Figure 2.1.1-1 will not affect the 
safety function and will not involve any change to the design or 
operation of any plant system or component. The revised Figure 2.1.1-1 
restricts reactor coolant flow to within previously analyzed 
temperature and pressure conditions. Therefore, no accident 
probabilities or consequences will be impacted.

Second Standard

    No. The proposed changes will not lead to any hardware or operating 
procedure change. Hence, no new equipment failure modes or accidents 
from those previously evaluated will be created.

Third Standard

    No. Margin of safety is associated with confidence in the design 
and operation of the plant; specifically, the ability of the fission 
product barriers to perform their design functions during and following 
an accident. The proposed changes to Figure 2.1.1-1 do not involve any 
change to plant design, operation, or analysis. Thus, the margin of 
safety previously analyzed and evaluated is maintained.
    Based on this analysis, it appears that the three standards of 10 
CFR 50.92(c) are satisfied for the proposed change to Figure 2.1.1-1. 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: June 24, 1999.
    Description of amendment request: The proposed amendments would 
change the Technical Specifications (TS) as follows: (1) Revise Figure 
2.1.1-1, ``Reactor Core Safety Limits--Four Loops in Operation,'' which 
defines the current limits of reactor coolant system (RCS) flow under 
different combinations of pressure and temperature; (2) revise the 
Actions associated with Limiting Condition of Operation (LCO) 3.4.1 and 
Table 3.4.1-1 to reflect the updated assumptions for reactor coolant 
flow, temperature and pressure; and (3) delete Figure 3.4.1-1, ``RCS 
Total Flow Rate Versus Rated Thermal Power--Four Loops in Operation,'' 
since these requirements are being relocated to LOC 3.4.1 and Table 
3.4.1-1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration for the June 24, 1999, submittal, which is presented 
below:

First Standard

    No component modification, system realignment, or change in 
operating procedure will occur which could affect the probability of 
any accident or transient. The increase in RCS total flow rate limit 
will not change the probability of actuation of any Engineered 
Safety Feature or other device. In order to provide more margin in 
the core design limits and allow more flexibility for future cycle-
specific core design, the analyses that establish these limits were 
reanalyzed at the proposed TS minimum RCS total flow rate limit. The 
impact of the power/flow tradeoff is determined for each reanalyzed 
event either by qualitative evaluation or by explicit reanalysis.
    An increase in the Technical Specification minimum RCS total 
flow rate limit and the revised power/flow tradeoff will not 
adversely affect the steady-state or transient analyses documented 
in Chapters 3, 4, 6, and 15 of the McGuire and Catawba Nuclear 
Station UFSARs [Updated Final Safety Analysis Reports]. The reduced 
RCS low flow reactor trip setpoint and allowable value will not 
increase the consequences of the partial loss of forced reactor 
coolant flow and reactor coolant pump shaft seizure accidents. In 
these transient reanalyses, the minimum DNBR and peak primary system 
pressure acceptance criteria are not adversely affected. Therefore, 
the proposed changes will not involve an increase in the probability 
or consequences of an accident previously evaluated.

Second Standard

    No component modification, system realignment, or change in 
operating procedure will occur which could create the possibility of 
a new or different kind of accident. As described in Attachment 3, 
the proposed increase in Technical Specification minimum RCS total 
flow rate limit and revised power/flow tradeoff will not adversely 
affect the steady-state or transient analyses documented in Chapters 
3, 4, 6, and 15 of the McGuire and Catawba Nuclear Station UFSARs. 
Therefore, the proposed changes will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.

Third Standard

    These amendments will not involve a significant reduction in a 
margin of safety. As described in Attachment 3, the increase in 
minimum RCS total flow rate limit and revised power/flow tradeoff 
will not adversely affect the steady-state or transient analyses 
documented in Chapters 3, 4, 6, and 15 of the McGuire and Catawba 
Nuclear Station UFSARs. DNBR, fuel clad intergrity, reactor vessel 
integrity and containment integrity will not be adversely affected 
by the proposed changes. Therefore, the proposed changes will not 
involve any reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina.
    NRC Section Chief: Richard L. Emch, Jr.

Duke Energy Corporation, et al., Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: July 22, 1998, supplemented by October 
22, 1998, January 28, May 6 and June 24, 1999.
    Description of amendment request: By the referenced submittals the 
licensee requested the McGuire Technical Specifications be changed to 
permit the licensee's planned use of fuel supplied by Westinghouse, 
which has different design characteristics from the fuel currently in 
use. The staff has previously published two Notices of

[[Page 43772]]

Consideration of Issuance of Amendments and Proposed No Significant 
Hazards Consideration of Issuance of Amendments. The first notice, 
dated December 16, 1998 (63 FR 69338), covers the submittals dated July 
22 and October 22, 1998. The second notice, dated May 19, 1999 (64 FR 
35202), covers the submittal dated May 6, 1999. The June 24, 1999, 
submittal actually requested an amendment separate from that described 
above, but nevertheless conveyed a revised proposed Figure 2.1.1-1, 
``Reactor Core Safety Limits--Four Loops in Operation,'' superseding 
what was originally proposed in the licensee's previous submittals. 
Hence this Notice only covers the revised proposed Figure 2.1.1-1. The 
Notices referenced above are unaffected.
    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 for the June 24, 1999, submittal. The staff has reviewed 
the licensee's analysis, and has performed its own analysis as follows:

First Standard

    No. The proposed changes to Figure 2.1.1-1 will not affect the 
safety function and will not involve any change to the design or 
operation of any plant system or component. The revised Figure 2.1.1-1 
restricts reactor coolant flow to within previously analyzed 
temperature and pressure conditions. Therefore, no accident 
probabilities or consequences will be impacted.

Second Standard

    No. The proposed changes would not lead to any hardware or 
operating procedure change. Hence, no new equipment failure modes or 
accidents from those previously evaluated will be created.

Third Standard

    No. Margin of safety is associated with confidence in the design 
and operation of the plant; specifically, the ability of the fission 
product barriers to perform their design functions during and following 
an accident. The proposed changes to Figure 2.1.1-1 do not involve any 
change to plant design, operation or analysis. Thus, the margin of 
safety previously analyzed and evaluated is maintained.
    Based on this analysis, it appears that the three standards of 10 
CFR 50.92(c) are satisfied for the proposed change to Figure 2.1.1-1. 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, Charlotte, North Carolina.
    Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina.
    NRC Section Chief: Richard L. Emch, Jr.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: June 24, 1999.
    Description of amendment request: The proposed amendments would 
change the Technical Specifications (TS) as follows: (1) Revise Figure 
2.1.1-1, ``Reactor Core Safety Limits--Four Loops in Operation,'' which 
defines the current limits of reactor coolant system (RCS) flow under 
different combinations of pressure and temperature; (2) revise Table 
3.3.1-1 to provide values for the trip setpoint and allowable value for 
RCS Flow-Low; (3) revise Table 3.3.1-1 to make a typographical 
correction for T, the nominal T-average at Rated Thermal Power; (4) 
revise the Actions associated with Limiting Condition of Operation 
(LCO) 3.4.1 and Table 3.4.1-1 to reflect the updated assumptions for 
reactor coolant flow, temperature and pressure; and (5) delete Figure 
3.4.1-1, ``RCS Total Flow Rate Versus Rated Thermal Power--Four Loops 
in Operation,'' since these requirements are being relocated to LCO 
3.4.1 and Table 3.4.1-1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

First Standard

    No component modification, system realignment, or change in 
operating procedure will occur which could affect the probability of 
any accident or transient. The increase in RCS total flow rate limit 
will not change the probability of actuation of any Engineered 
Safety Feature or other device. In order to provide more margin in 
the core design limits and allow more flexibility for future cycle-
specific core design, the analyses that establish these limits were 
reanalyzed at the proposed TS minimum RCS total flow rate limit. The 
impact of the power/flow tradeoff is determined for each reanalyzed 
event either by qualitative evaluation or by explicit reanalysis.
    An increase in the Technical Specification minimum RCS total 
flow rate limit and the revised power/flow tradeoff will not 
adversely affect the steady-state or transient analyses documented 
in Chapters 3, 4, 6, and 15 of the McGuire and Catawba Nuclear 
Station UFSARs [Updated Final Safety Analysis Reports]. The reduced 
RCS low flow reactor trip setpoint and allowable value will not 
increase the consequences of the partial loss of forced reactor 
coolant flow and reactor coolant pump shaft seizure accidents. In 
these transient reanalyses, the minimum DNBR and peak primary system 
pressure acceptance criteria are not adversely affected. Therefore, 
the proposed changes will not involve an increase in the probability 
or consequences of an accident previously evaluated.

Second Standard

    No component modification, system realignment, or change in 
operating procedure will occur which could create the possibility of 
a new or different kind of accident. As described in Attachment 3, 
the proposed increase in Technical Specification minimum RCS total 
flow rate limit and revised power/flow tradeoff will not adversely 
affect the steady-state or transient analyses documented in Chapters 
3, 4, 6, and 15 of the McGuire and Catawba Nuclear Station UFSARs. 
Therefore, the proposed changes will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.

Third Standard

    These amendments will not involve a significant reduction in a 
margin of safety. As described in Attachment 3, the increase in 
minimum RCS total flow rate limit and revised power/flow tradeoff 
will not adversely affect the steady-state or transient analyses 
documented in Chapters 3, 4, 6, and 15 of the McGuire and Catawba 
Nuclear Station UFSARs. DNBR, fuel clad integrity, reactor vessel 
integrity and containment integrity will not be adversely affected 
by the proposed changes. Therefore, the proposed changes will not 
involve any reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, Charlotte, North Carolina.
    Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina.
    NRC Section Chief: Richard L. Emch, Jr.

[[Page 43773]]

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

    Date of amendment request: July 14, 1999.
    Description of amendment request: The proposed amendments delete 
requirements from the Technical Specifications to maintain a Post 
Accident Sampling System (PASS). Licensees were required to implement 
PASS upgrades as a result of NUREG-0737, ``Clarification of TMI [Three 
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97, 
Revision 3, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Access Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades were an outcome of the 
NRC's lessons learned from the accident that occurred at Three Mile 
Island, Unit 2. EOI has stated that the information obtained using PASS 
can be readily obtained through other means or is of little use in the 
assessment and mitigation of accident conditions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1--[The Proposed Change] Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated

    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the Three Mile Island Unit 2 (TMI-2) accident. The specific 
intent of the PASS was to provide a system that has the capability 
to obtain and analyze samples of plant fluids containing potentially 
high levels of radioactivity, without exceeding plant personnel 
radiation exposure limits. Analytical results of these samples would 
be used largely for verification purposes in aiding the plant staff 
in assessing the extent of core damage and subsequent offsite 
radiological dose projections.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that the actual benefits afforded by a 
PASS provide little benefit to post accident mitigation. Past 
experience has indicated that there exists in-plant instrumentation 
and methodologies available in lieu of a PASS for collecting and 
assimilating information needed to assess core damage following an 
accident. Furthermore, the implementation of Severe Accident 
Management Guidance (SAMG) emphasizes accident management strategies 
based on in-plant instruments. These strategies provide guidance to 
the plant staff for mitigation and recovery from a severe accident. 
Based on current severe accident management strategies and 
guidelines, it is determined that the PASS provides no benefit to 
the plant staff in coping with an accident. The use of the PASS may 
be counter productive to plant operations since its operation will 
divert resources away from accident management, the sample results 
may be ambiguous and may be misinterpreted, and the use of PASS may 
restrict personnel movements in certain areas of the plant while 
resulting in additional fission product release points outside the 
containment.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Additionally, preliminary discussions with the State of Arkansas 
have indicated that the elimination of the PASS will not adversely 
impact actions taken by the State during an emergency event. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 [accident] 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan PARs [protective action 
recommendations].
    Therefore, the elimination of PASS requirements of the ANO-1 and 
ANO-2 [Arkansas Nuclear One, Unit 1 and Unit 2] Technical 
Specifications (TS) and subsequent requested relief from the 
requirements of NUREG-0737 and Regulatory Guide 1.97, Revision 3, 
does not involve a significant increase in the probability or 
consequences of any accident previously evaluated.

Criterion 2--[The Proposed Change] Does Not Create the Possibility 
of a New or Different Kind of Accident from any Previously 
Evaluated

    The relief from PASS related NUREG-0737 and Regulatory Guide 
1.97 requirements in addition to the proposed TS changes will not 
result in any failure mode not previously analyzed. The PASS was 
intended to allow for verification of the extent of reactor core 
damage and also to provide an input to offsite dose projection 
calculations. The PASS is not considered an accident precursor, nor 
does its existence or elimination have any adverse impact on the 
pre-accident state of the reactor core or post accident confinement 
of radionuclides within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--[The Proposed Change] Does Not Involve a Significant 
Reduction in the Margin of Safety

    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety at ANO-1 and ANO-2. Non-
PASS methodologies are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events nor rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Therefore, based upon the reasoning presented above and the 
previous discussion of the amendment request, Entergy Operations, 
Inc. has determined that the requested change does not involve a 
significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Indiana Michigan Power Company, Docket Nos. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan.

    Date of amendment requests: December 3, 1998.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification (TS)
3/4.7.7, ``Sealed Source Contamination,'' and the associated bases to 
address testing requirements for fission detectors. The proposed 
changes would provide consistency between the unit 1 and Unit 2 TS 
requirements and with NUREG-0452, ``Standard Technical 
Specifications.''.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

[[Page 43774]]

Criterion 1

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes clarify testing requirements for fission 
detectors. When the fission detectors are tested for surface 
contamination, they do not interfere with plant equipment and they 
do not affect plant operation. The detectors are not assumed to 
initiate an accident; therefore, the probability of an accident 
previously evaluated is not changed.
    Conducting tests prior to using a new fission detector provides 
assurance that intake limits will not be exceeded. There is no 
change to the nuclear material contained in the detector. The 
fission detectors are not used to mitigate the consequences of 
postulated accidents. Therefore, the consequences of an accident 
remain the same as previously evaluated.
    Therefore, it is concluded that the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.

Criterion 2

    Does the change create the possibility of a new or different 
type of accident from any accident previously evaluated?
    The proposed changes do not affect the design or operation of 
systems, structures, or components in the plant. There are no 
changes to parameters governing plant operation, and no new or 
different types of equipment will be installed. Therefore, it is 
concluded that the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.

Criterion 3

    Does the change involve a significant reduction in a margin of 
safety?
    The proposed changes do not introduce new equipment, equipment 
modifications, or new or different modes of plant operation. These 
changes do not affect the operational characteristics of any 
equipment or systems.
    Therefore, it is concluded that these changes do not involve a 
significant reduction in the margin of safety.

Conclusion

    In summary, based upon the above evaluation, the Licensee has 
concluded that these changes involve no significant hazards 
consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, MI 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

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

    Date of application for amendments: March 29, 1999.
    Description of amendment request: The proposed amendments would 
revise Technical Specification Surveillance Requirement 3.9.1.1 and the 
associated Bases 3.9.1 to delete the requirement for the refuel 
platform fuel grapple fully retracted position interlock.
    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 TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This change removes a redundant interlock and will not impact 
the functionality of associated interlocks. The removal of the 
``refuel platform fuel grapple fully retracted position'' refueling 
interlock will not affect the ability of the remaining refueling 
interlocks to produce a rod block during fuel moves. The 
administrative controls in place do not allow control rod 
withdrawals while fuel is being moved or fuel movement while rods 
are withdrawn. The fuel grapple full up interlock is a redundant and 
diverse interlock and its removal has no impact on plant safety. The 
interlock's intent, to provide a backup to the load sensor, is not 
required since the setpoint is currently low enough to provide 
adequate protection therefore not significantly increasing the 
probability of an accident previously evaluated.
    The refueling interlocks are not used to prevent or to mitigate 
the fuel handling accident as discussed in the PBAPS [Peach Bottom 
Atomic Power Station], Units 2 and 3, UFSAR [Updated Final Safety 
Analysis Report], Section 14.6.4 (``Refueling Accident''). The 
``refuel platform fuel grapple fully retracted position'' interlock 
and the ``refuel platform fuel grapple, fuel loaded'' interlock both 
provide rod blocks during fuel movement over the core. Additionally, 
the refueling interlocks are not assumed as an initial condition in 
the control rod drop accident as discussed in the PBAPS, UFSAR, 
Section 14.6.2 (``Control Rod Drop Accident''). The control rod drop 
accident is only analyzed when the reactor is critical and not 
during refueling operations.
    The refueling interlocks associated with the refueling platform 
provide rod blocks to ensure that control rods can not be withdrawn 
when fuel is being moved over the core (PBAPS, Units 2 and 3, UFSAR 
Section 14.5.3.3, ``Control Rod Removal Error During Refueling''). 
They are also used to prevent refueling bridge motion towards the 
core if a control rod is withdrawn during fuel movements (PBAPS, 
Units 2 and 3, UFSAR Section 14.5.3.4, ``Fuel Assembly Insertion 
Error During Refueling''). These interlocks prevent the possibility 
of an inadvertent criticality during refueling. However, removal of 
the ``refuel platform fuel grapple fully retracted position'' 
interlock, which is a redundant and diverse interlock, will not 
prevent the remaining interlocks from performing their intended 
safety functions. The refueling interlocks are active with the mode 
switch in refuel, and are only designed to reinforce administrative 
procedures for moving fuel. Therefore, the proposed TS changes will 
not involve a significant increase in the probability of an accident 
previously evaluated.
    The fuel or core loading characteristics are not altered by the 
removal of this interlock. The dose resulting from a potential 
control rod withdrawal or fuel bundle error event is not increased 
as a result of eliminating this redundant and diverse interlock. 
Therefore, the removal of the ``refuel platform fuel grapple fully 
retracted position'' interlock will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The refueling interlocks are not accident initiators. Nor will 
any new failure mode be introduced by the removal of the ``refuel 
platform fuel grapple fully retracted position'' interlock. The 
interlocks are used to reinforce administrative controls which 
prevent fuel movement over the core with control rods withdrawn and 
preclude withdrawal of control rods when the fuel is being moved 
over the core. The interlock for ensuring the fuel grapple is fully 
up, is a redundant and diverse interlock since a load sensor 
determines if the main hoist is loaded with a fuel bundle. This 
redundant and diverse interlock prevents the withdrawal of a control 
rod while moving fuel during refueling. The setpoint is low enough 
to ensure a rod block will be received if the main hoist is being 
used to move fuel over the core and to prevent movement of the 
refueling bridge. The remaining refueling interlocks, in combination 
with the refueling procedures, will still prevent an inadvertent 
criticality during refueling operations. Fuel handling procedures 
require that interlocks be verified by observing the rod withdraw 
permissive light in the control room, and by monitoring the rod 
block interlock light on the refuel bridge. Therefore, the proposed 
TS changes do not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    This change will not involve a significant reduction in a margin 
of safety. The ``refuel platform fuel grapple fully retracted 
position'' interlock is redundant and diverse to the ``refuel 
platform fuel grapple, fuel

[[Page 43775]]

loaded'' interlock on the main hoist. The other two hoists on the 
bridge have the fuel loaded interlock but do not have the backup 
full up position interlock. The margin of safety of the refueling 
interlocks will not be significantly reduced by this change since 
redundant interlocks are not required (this a nonsafety-related 
function) and the original justification for using it, a high load 
weight setpoint, is no longer applicable. The system consists of a 
single channel, and no current design basis for using redundant and 
diverse interlocks to provide the rod block. Additionally, the 
Reactor Manual Control System will not be affected by this change. 
The system's ability to provide a rod block is not affected by this 
change.

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

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

    Date of amendment request: April 5, 1999
    Description of amendment request: The proposed changes would revise 
Appendix A (Section 6.1) and Appendix B (Section 7.1) of the James A. 
FitzPatrick Technical Specifications. The proposed changes would remove 
the position title of General Manager from these sections and would 
state that if the Site Executive Officer (SEO) is unavailable, he will 
delegate his responsibilities to another staff member, in writing. In 
addition the position title of Resident Manager, used in Apendix B, 
Section 7.1, would be replaced by Site Executive Officer.
    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:

    Consistent with the criteria of 10 CFR 50.92, the proposed 
application is judged to involve no significant hazards based on the 
following information:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously analyzed?
    Response: The proposed changes to Appendix A (Section 6.1) and 
Appendix B (Section 7.1) are administrative in nature in that they 
do not change the intent of the Technical Specifications. If the SEO 
is unavailable, he will still delegate his responsibilities to a 
qualified personnel member, such as the Plant Manager or one of the 
General Managers. These changes can not cause an accident or 
contribute to the probability or consequences of one.
    The replacement of the position title of Resident Manager with 
Site Executive Officer in Appendix B, Section 7.1, was already 
approved by the NRC in Amendment 228.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously analyzed.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response: The proposed changes to Appendix A (Section 6.1) and 
Appendix B (Section 7.1) are administrative in nature as they do not 
affect the function of plant equipment or the way the equipment 
operates. The changes do not change the intent of the current TS, in 
that if the SEO is unavailable, he will delegate his 
responsibilities to another personnel member such as the Plant 
Manager or one of the General Managers. Appendix A (Section 6.1) and 
Appendix B (Section 7.1) are being revised to eliminate the need for 
future TS changes to these sections resulting solely from the 
creation of new or revised management positions (such as the Plant 
Manager), title changes to the position of General Manager, or a 
change to the number of General Managers. These types of 
organizational changes will be evaluated using the criteria of 10 
CFR 50.59.
    The replacement of the position title of Resident Manager with 
Site Executive Officer in Appendix B, Section 7.1, was already 
approved by the NRC in Amendment 228.
    Therefore, the proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: The proposed changes to Appendix A (Section 6.1) and 
Appendix B (Section 7.1) are administrative changes associated with 
the delegation of the SEO's responsibilities when he is unavailable. 
These changes do not change the intent of the current TS, in that in 
the SEO's absence, he will still delegate his responsibilities to 
other personnel members such as the Plant Manager or General 
Managers.
    The replacement of the position title of Resident Manager with 
Site Executive Officer in Appendix B, Section 7.1, was already 
approved by the NRC in Amendment 228.
    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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: S. Singh Bajwa.

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

    Date of amendment request: June 22, 1999.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications by changes to the Pressure and 
Temperature (P-T) limits. As part of this proposed change the licensee 
is proposing to add separate bottom head curves ABH and 
BBH for in-service hydrostatic and leak tests and non-
nuclear heatup and cooldown, respectively. In addition, a non-beltline 
curve (i.e., ANB) for in-service hydrostatic and leak tests 
is being proposed.
    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:

    Operation of the FitzPatrick plant in accordance with the 
proposed amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The changes to the P-T curves are being proposed to preclude 
brittle fracture of RPV [Reactor Pressure Vessel] materials for up 
to 32 EFPY [effective full-power years]. In addition to the P-T 
curve for up to 32 EFPY, a P-T curve has been prepared for exposures 
up to 24 EFPY to shorten outage time for startups conducted prior to 
reaching this exposure. Safety margins specified in 10 CFR 50, 
Appendix G and Appendix G to Section XI of the ASME [American 
Society of Mechanical Engineers Boiler and Pressure Vessel Code] 
will continue to be met for each of these curves. Therefore, there 
is not a significant increase in the probability of an accident 
previously evaluated.
    The RPV, as part of the reactor coolant system, provides a 
barrier to the release of reactor coolant. Operation in accordance

[[Page 43776]]

with the proposed amendment will preclude brittle fracture of the 
RPV consistent with current requirements, and consequently, does not 
significantly increase the consequences of an accident previously 
evaluated.
    Based on the above, operation of the FitzPatrick plant in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not involve any physical alterations to 
plant configurations or introduce any new accident precursors which 
could initiate a new or different kind of accident. The proposed 
change does not affect the intended function of the RPV nor does it 
affect the operation of the RPV in a way which would create a new or 
different kind of accident. The changes to the P-T curves are being 
proposed to preclude brittle fracture of RPV materials for up to 32 
EFPY. Safety margins specified in 10 CFR 50, Appendix G and Appendix 
G to Section XI of the ASME Code will continue to be met. Therefore, 
operation of the FitzPatrick plant in accordance with the proposed 
amendment will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The existing FitzPatrick P-T curves were developed using safety 
margins for brittle fracture found in 10 CFR 50 Appendix G. The 
proposed FitzPatrick P-T curves, which are valid for up to 32 EFPY 
of operation, were also developed using safety margins for brittle 
fracture found in 10 CFR 50 Appendix G. Based on this, operation of 
the FitzPatrick plant in accordance with the proposed amendment will 
continue to preclude brittle fracture of the RPV materials during 
in-service hydrostatic and leak tests, non-nuclear heatup and 
cooldown, and core critical operation without a significant 
reduction in a margin of safety. Therefore, operation of the 
FitzPatrick plant in accordance with the proposed amendment will not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: S. Singh Bajwa.

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

    Date of amendment request: July 2, 1999.
    Description of amendment request: The proposed amendments would 
relocate the requirements from Technical Specification 3/4.3.4, 
``Instrumentation, Turbine Overspeed Protection,'' and the associated 
bases to licensee-controlled documents in accordance with Generic 
Letter 95-10, ``Relocation of Selected Technical Specifications 
Requirements Related to Instrumentation.''
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The requested amendments will not involve an increase in the 
probability or consequences of an accident previously evaluated. 
Relocation of the affected Technical Specification sections and 
their Bases to the Salem UFSAR [Updated Final Safety Analysis 
Report] will have no affect on the probability that any accident 
will occur. Additionally, the consequences of an accident will not 
be affected because the Turbine Overspeed Protection system will 
continue to be utilized in the same manner as before. No impact on 
the plant response to accidents will be created.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed amendments will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. No new accident causal mechanisms will be created as a 
result of the relocation of the Turbine Overspeed Protection system 
Technical Specification requirements and their Bases to the Salem 
UFSAR. Plant operation will not be affect by the proposed amendments 
and no new failure modes will be created.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed amendments will not involve a reduction in the 
margin of safety. Relocation of the affected Technical Specification 
requirements to the Salem UFSAR is consistent with NUREG 1431, 
Standard Technical Specifications--Westinghouse Plants which do not 
include Technical Specification requirements for the Turbine 
Overspeed Protection system. The proposed amendments are consistent 
with the NRC philosophy of encouraging utilities to propose 
amendments that are consistent with NUREG 1431.
    Based on the above, the proposed changes will not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: July 16, 1999.
    Description of amendment request: Proposed Technical Specifications 
(TS) change to increase the action requirement time to be in Mode 3 if 
the temperature of the ultimate heat sink (UHS) exceeds the TS limit of 
75  deg.F. The increased time will only apply if the UHS temperature is 
between 75 and 77  deg.F. The Bases for the associated TS will also be 
revised.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes will allow plant operation to continue for 
an additional 12 hours with the temperature of the Ultimate Heat 
Sink (UHS) up to 2  deg.F above the Technical Specification limit of 
75  deg.F. This increase in UHS temperature will not affect the 
normal operation of the plant to the extent which would make any 
accident more likely to occur. In addition, there exists adequate 
margin in the safety systems and heat exchangers to assure the 
safety functions are met at the higher temperature. An evaluation 
has confirmed that safe shutdown will be achieved and maintained for 
a loss of coolant accident (LOCA) with a loss of normal power (LNP) 
and a single active failure with a UHS water temperature as high as 
77  deg.F.
    The proposed changes will have no adverse effect on plant 
operation, or the availability or operation of any accident 
mitigation equipment. The plant response to the design basis 
accidents will not change. In addition, the proposed changes can not 
cause an accident. Therefore, there will be no significant increase 
in the probability or consequences of an accident previously 
evaluated.

[[Page 43777]]

    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes will allow plant operation to continue for 
an additional 12 hours with the temperature of the UHS up to 2 
deg.F above the Technical Specification limit of 75  deg.F. This 
will not alter the plant configuration (no new or different type of 
equipment will be installed) or require any new or unusual operator 
actions. The proposed changes will not alter the way any structure, 
system, or component functions and will not significantly alter the 
manner in which the plant is operated. There will be no adverse 
effect on plant operation or accident mitigation equipment. The 
proposed changes do not introduce any new failure modes. Also, the 
response of the plant and the operators following these accidents is 
unaffected by the changes. In addition, the UHS is not an accident 
initiator. Therefore, the proposed changes will not create the 
possibility of a new or different kind of accident from any 
previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes will allow plant operation to continue for 
an additional 12 hours with the temperature of the UHS up to 2 
deg.F above the Technical Specification limit of 75  deg.F. 
Evaluations have been performed which demonstrate that the safety 
systems have adequate margin to ensure their safety functions can be 
met with a UHS temperature of 77  deg.F. In addition, safe shutdown 
capability has been demonstrated for a UHS water temperature as high 
as 77  deg.F.
    The proposed changes will have no adverse effect on plant 
operation or equipment important to safety. The plant response to 
the design basis accidents will not change and the accident 
mitigation equipment will continue to function as assumed in the 
design basis accident analysis. Therefore, there will be no 
significant reduction in a margin of safety.

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

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Docket No. 
50-278, Peach Bottom Atomic Power Station, Unit No. 3, York County, 
Pennsylvania

    Date of application for amendment: July 12, 1999.
    Description of amendment request: The proposed change will revise 
Technical Specifications (TSs) TS 2.1.1.2, ``Reactor Core [Safety 
Limits] SLs,'' and Section 5.6.5, ``Core Operating Limits Report.'' 
These Sections will be revised to: (1) Incorporate revised Safety Limit 
Minimum Critical Power Ratios (SLMCPRs) due to the use of a cycle-
specific analysis performed by General Electric Nuclear Energy (GENE) 
for Peach Bottom Atomic Power Station, Unit 3, (PBAPS, Unit 3) Cycle 
13, (2) delete previously added footnotes which are no longer 
necessary, and (3) update a reference contained in TS 5.6.5.b.2 which 
documents an analytical method used to determine the core operating 
limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The derivation of the cycle specific SLMCPRs for incorporation 
into the TS, and its use to determine cycle specific thermal limits, 
has been performed using the methodology discussed in ``General 
Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13, 
and U.S. Supplement, NEDE-24011-P-A-13-US, August 1996, and 
Amendment 25. Amendment 25 was approved by the NRC in a March 11, 
1999 safety evaluation report. This change in SLMCPRs cannot 
increase the probability or severity of an accident.
    The basis of the SLMCPR calculation is to ensure that greater 
than 99.9% of all fuel rods in the core avoid transition boiling if 
the limit is not violated. The new SLMCPRs preserve the existing 
margin to transition boiling and fuel damage in the event of a 
postulated accident. The fuel licensing acceptance criteria for the 
SLMCPR calculation apply to PBAPS, Unit 3, Cycle 13 in the same 
manner as they have applied previously. The probability of fuel 
damage is not increased. Therefore, the proposed TS changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    In addition to the change to the SLMCPR, the footnotes to TS 
2.1.1.2 and TS 5.6.5.b.1 are being deleted. The footnote associated 
with TS 2.1.1.2 was originally included to ensure that the SLMCPR 
value was only applicable for the identified cycle. The footnote was 
added to TS 5.6.5.b.1 because Amendment 25 and the R-factor 
calculation methodology were not yet NRC approved. Amendment 25 and 
the R-factor methodology have subsequently been approved. Therefore, 
these footnotes are no longer necessary. The footnotes were for 
information only, and have no impact on the design or operation of 
the plant. The deletion of the footnotes associated with TS 2.1.1.2 
and TS 5.6.5.b.1 is an administrative change that does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The Revision 1 ARTS/MELLLA [Maximum Extended Load Line Limit and 
ARTS Improvement Program Analysis for Peach Bottom Atomic Power 
Station Unit 2 and 3,] analysis contained in TS 5.6.5.b.2 is being 
updated to a Revision 2 analysis, to reflect changes that were 
previously approved by the NRC as documented in the safety 
evaluation report dated August 10, 1994 (Amendment No. 192 for 
PBAPS, Unit 2). This is an administrative change which will ensure 
that the references contained in the PBAPS Technical Specifications 
are accurate and consistent with other licensing documents. No 
technical changes are occurring which have not been previously 
approved by the NRC. Therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The SLMCPR is a TS numerical value, designed to ensure that 
transition boiling does not occur in 99.9% of all fuel rods in the 
core during the limiting postulated accident. The new SLMCPRs are 
calculated using NRC approved methodology discussed in ``General 
Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13 
(GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-13-US, August 1996, 
and Amendment 25. The SLMCPR is not an accident initiator, and its 
revision will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    Additionally, this proposed change will delete footnotes 
contained in TS 2.1.1.2 and TS 5.6.5.b.1 as the result of the NRC 
approval of analysis associated with Amendment 25 and the R-factor 
methodology. The proposed change also updates the ARTS/MELLLA 
analysis contained in TS 5.6.5.b.2. This revision contains 
information which was previously approved by the NRC. Therefore, the 
deletion of the footnotes associated with TS 2.1.1.2 and TS 
5.6.5.b.1, and the updating of the reference contained in TS 
5.6.5.b.2 are administrative changes that do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    There is no significant reduction in the margin of safety 
previously approved by the NRC as a result of: (1) the proposed 
changes

[[Page 43778]]

to the SLMCPRs, (2) the proposed change that will delete the 
footnotes to TS 2.1.1.2 and TS 5.6.5.b.1, and (3) updating the 
reference to the ARTS/MELLLA analysis contained in TS 5.6.5.b.2. The 
new SLMCPRs are calculated using methodology discussed in ``General 
Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13 
(GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-13-US, August 1996, 
and Amendment 25. The fuel licensing acceptance criteria for the 
calculation of the SLMCPR apply to PBAPS, Unit 3 Cycle 13 in the 
same manner as they have applied previously. The SLMCPRs ensure that 
greater than 99.9% of all fuel rods in the core will avoid 
transition boiling if the limit is not violated when all 
uncertainties are considered, thereby preserving the fuel cladding 
integrity. Therefore, the proposed TS changes will not involve a 
significant reduction in the margin of safety previously approved by 
the NRC.
    Additionally, the proposed changes that delete the footnotes to 
TS 2.1.1.2 and TS 5.6.5.b.1, and update the revision to the ARTS/
MELLLA analysis contained in TS 5.6.5.b.2, are administrative 
changes that will not significantly reduce the margin of safety 
previously approved by the NRC.

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

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

    Date of amendment request: June 28, 1999.
    Description of amendment request: The proposed amendment would 
revise the Improved Technical Specifications (ITS) associated with the 
Reactor Coolant System (RCS) Leakage Detection Instrumentation.
    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 Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. The changes add 
further requirements for redundancy and a requirement to perform 
either an RCS water inventory balance or analyses of containment 
atmosphere grab samples once within 12 hours and every 12 hours 
thereafter when the particulate containment atmosphere radioactivity 
monitor is unavailable while in Modes 1, 2, 3, and 4. This does not 
increase the probability of an accident previously evaluated since 
the compensatory actions are either a calculation utilizing 
installed indication or the measurement of a sample drawn downstream 
from the containment atmosphere sample isolation valves and are of 
themselves not an accident initiator. The proposed compensatory 
actions are based on the NUREG-1431 guidance and the proposed 
frequencies are more conservative, which gives a higher assurance 
that the RCS leakage rate can be adequately monitored.
    Therefore, the probability or consequences of an accident 
previously evaluated is not significantly increased.
    2. Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
changes add further requirements for redundancy and the proposed 
change for compensatory actions when the particulate containment 
atmosphere radioactivity monitor is inoperable does not of itself 
involve a physical alteration of the plant (ie. no new or different 
type of equipment will be added to perform the required actions) or 
changes in the methods governing normal plant operation. The changes 
only involve implementing currently approved alternate methods to 
determine the RCS leak rate on an increased frequency. Therefore, 
the possibility for a new or different kind of accident from any 
accident previously evaluated is not created.
    3. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. The proposed changes only add conservatism in the number of 
required RCS leakage detection instrumentation and add more 
conservative compensatory actions that are to be taken when the 
containment atmosphere particulate radioactivity monitor is 
inoperable. The compensatory actions are based on the guidance of 
NUREG-1431. Therefore, this change does not involve a significant 
reduction in a margin of safety.

    Based upon the preceding information, it has been determined that 
the proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated, create 
the possibility of a new or different kind of accident from any 
accident previously evaluated, or involve a significant reduction in a 
margin of safety. Therefore, it is concluded that the proposed changes 
meets the requirements of 10 CFR 50.92(c) and do not involve a 
significant hazards consideration.
    Local Public Document Room Location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610
    Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW, Washington, DC 20005.
    NRC Section Chief: S. Singh Bajwa.

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: September 10, 1998 (PCN-496), as 
supplemented July 19, 1999.
    Description of amendment requests: The proposed amendments would 
modify the Technical Specifications for the San Onofre Nuclear 
Generating Station (SONGS) Units 2 and 3 to delete the requirements for 
equipment used to control hydrogen in the containment structure.
    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
    The containment hydrogen control system is currently classified 
as an engineered safety feature that serves as the combustible gas 
control system in the containment. The hydrogen control system is 
composed of a hydrogen recombiner subsystem and a hydrogen purge 
subsystem. Hydrogen control subsystem components are not considered 
to be accident initiators.
    Therefore, this change does not increase the probability of an 
accident previously evaluated.
    The hydrogen control system is provided to ensure that the 
hydrogen concentration is maintained below the flammability limit of 
4% so that containment integrity is not challenged following a 
design basis Loss Of Coolant Accident (LOCA). Existing analysis 
show[s] that the hydrogen concentration will not reach the 
flammability limit of 4% for at least 13.5 days after a design basis 
LOCA. The time available will be extended to over 30 days using more 
realistic hydrogen generation rates. The containment peak pressure 
will remain below the San Onofre Nuclear Generating Station Units 2 
and 3 (SONGS 2 & 3) containment design pressure of 60 psig [pounds 
per square inch gauge] during this time. Beyond 30 days, hydrogen 
concentration may reach the flammability limit. However, containment 
failure due to hydrogen combustion is unlikely based on the results 
of the SONGS 2 & 3 IPE [indvidual plant examination] study. The 
detailed

[[Page 43779]]

SONGS 2 & 3-specific containment integrity analysis indicates that 
containment rupture pressure is approximately 139 psig with 95% 
confidence. Therefore, this change does not increase the 
consequences of accidents previously evaluated.
    Removal of the existing requirements for hydrogen control will 
eliminate the Emergency Operating Instruction (EOI) steps for 
hydrogen control and hence simplify the EOls. This would have a 
positive impact on public health risk by reducing the probability of 
operator error during potential accidents and hence reduce the core 
damage frequency. As proposed in this change request, these changes 
will allow the operators to address all hydrogen control issues as 
part of the proposed Accident Management Guidelines which cover 
operator actions at long time frames following accidents.
    Removal of the existing requirements for hydrogen control will 
eliminate the EOI steps to initiate the containment hydrogen purge. 
This will result in a lower probability of a failed open containment 
purge valve. Consequently, the offsite doses would be reduced due to 
the reduction of the probability of a failed-open containment purge 
valve. The changes described in this request result in a ``risk 
positive'' change.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any 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 or configuration 
of the plant beyond the hydrogen control system. Hydrogen generation 
following a design basis LOCA has been evaluated in accordance with 
regulatory requirements. Deletion of the hydrogen control system 
from the Technical Specifications does not alter the hydrogen 
generation processes post-LOCA. The consideration of hydrogen 
generation will no longer be included in the design basis of SONGS 2 
& 3. Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No
    The changes described in this change request result in a ``risk 
positive'' change. Removal of the existing requirement for a 
hydrogen control system will, by eliminating the EOI steps for 
hydrogen control, result in lower operator error probabilities. 
Elimination of the EOI steps to initiate the containment hydrogen 
purge will result in a lower probability of a failed-open 
containment purge valve, resulting in lower large early release 
probabilities.
    Therefore, this change involves an increase in safety, not a 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, Irvine, California 92713.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: April 13, 1999.
    Description of amendment request: Southern Nuclear Operating 
Company (SNC) proposes to revise the Vogtle Electric Generating Plant 
(VEGP) Unit 1 and Unit 2 Technical Specifications (TS) Limiting 
Condition for Operation (LCO) Applicability LCO 3.0.4 and Surveillance 
Requirement (SR) Applicability SR 3.0.4. The proposed changes would 
update the versions of LCO 3.0.4 and SR 3.0.4 that appear in the 
existing VEGP TS to be consistent with the versions of LCO 3.0.4 and SR 
3.0.4 as they appear in Revision 1 to NUREG-1431. The proposed change 
would add the words ``or that are part of a shutdown of the unit,'' to 
LCO 3.0.4 to allow reactor shutdowns that are not necessarily required 
by other TS Required Actions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed change has impact on what equipment is required 
to be OPERABLE or demonstrated OPERABLE via surveillance prior to 
unit shutdowns or entry into MODES 5 and 6. This change could 
increase the probability or consequences of an accident previously 
evaluated if applied without consideration to all applicable 
transitions. However, as part of the change, an evaluation is 
attached in the form of a matrix that identifies those 
specifications to which LCO 3.0.4 and SR 3.0.4 must continue to 
apply. Therefore, only those specifications that do not impact 
safety for these plant conditions are afforded this relaxation. As 
such, there is no increase in the probability or consequences of an 
accident previously evaluated as this assessment has been performed 
and documented with the submittal.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed change administratively changes when equipment 
is required to be OPERABLE or demonstrated OPERABLE via surveillance 
prior to unit shutdown or entry into MODES 5 and 6. However, as no 
changes in equipment function or operation are included, there is no 
increase in the probability of a new or different kind of accident 
from those previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    No. The proposed change has impact on what equipment is required 
to be OPERABLE or demonstrated OPERABLE via surveillance prior to 
unit shutdown or entry into MODES 5 and 6. This change could impact 
the margin of safety of some accidents if applied without 
consideration to all applicable transitions. However, as part of the 
change, an evaluation is attached in the form of a matrix, that 
identifies those specifications to which LCO 3.0.4 and SR 3.0.4 must 
continue to apply. Therefore, only those specifications that do not 
impact safety for these plant conditions, which includes any impact 
on margin of safety are afforded this relaxation. As such, there is 
no reduction in the margin of safety.

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

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

    Date of amendment request: April 28, 1999.
    Description of amendment request: The amendments revise Vogtle's 
licensing basis to allow the licensee to establish containment hydrogen 
monitoring within 90 minutes of initiation of a safety injection 
following a loss-of-coolant accident, compared to the current 30 
minutes requirement.
    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

[[Page 43780]]

consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Containment hydrogen concentration is not an input parameter to 
the FSAR Chapter 15 accident analyses for a loss of reactor or 
secondary coolant accidents; nor is it used as an initial assumption 
for the containment response analysis. Control room operators use 
the containment hydrogen monitors to establish hydrogen control 
measures should it become necessary. However, the actions required 
to establish containment hydrogen monitoring are a distraction for 
the operators from more important tasks during the early phases of 
an accident. Hydrogen production occurs over a long period and a 
significant accumulation is not expected for several hours into the 
event. This function is more appropriately included as a part of the 
long-term core damage assessment process. The one-hour extension 
will have a positive impact on the ability of the operators to 
concentrate on their more immediate actions while having no negative 
impact on the long-term assessment efforts. Therefore, the proposed 
license amendment 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 
amendment would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    Operation of the containment hydrogen monitors is not an 
initiator of any design basis accident. Control room operators use 
the containment hydrogen monitors following a LOCA to establish 
hydrogen control measures should it become necessary. Accurate 
indication of containment hydrogen concentration is needed prior to 
initiating recombiner operation or containment venting and for long-
term core damage assessment. The proposed license amendment would 
not eliminate the requirement to establish hydrogen monitoring, but 
would permit it to be delayed until those actions required to 
diagnose the event and verify proper operation of essential safety 
equipment have been completed. The one-hour extension maintains the 
requirement to establish hydrogen monitoring well before calculated 
conditions inside the containment indicate any need to initiate 
hydrogen control measures. Therefore, the proposed license amendment 
will not create a new or different kind of accident from any 
previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The need to establish hydrogen control measures will not be 
present within the first 90 minutes following a LOCA since there 
will not be significant hydrogen accumulation. By extending the time 
allowed to establish containment hydrogen monitoring, the operators 
can remain focused on the actions necessary to assess and mitigate 
the accident before redirecting their attention to long-term 
recovery actions. The one-hour extension maintains the requirement 
to establish hydrogen monitoring well before calculated conditions 
inside the containment indicate any need to initiate hydrogen 
control measures. Therefore, the proposed license amendment will not 
involve a significant reduction in a margin of safety, but will 
instead result in an overall enhancement to safety.

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

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

    Date of amendment request: May 18, 1999.
    Description of amendment request: The proposed change would revise 
Surveillance Requirements (SRs) 3.8.1.3 and 3.8.1.13 to reduce the 
loading requirements for the diesel generators (DGs). Presently, SR 
3.8.1.3 requires that the DGs be loaded and operated for greater than 
or equal to 60 minutes between 6800 kW and 7000 kW at least once every 
31 days. The proposed change would revise the lower end of the load 
band in SR 3.8.1.3 to 6500 kW from 6800 kW. Revised SR 3.8.1.3 would 
require that the DGs be loaded and operated for greater than or equal 
to 60 minutes at a load greater than or equal to 6500 kW and less than 
or equal to 7000 kW at least once every 31 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. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed change affects only the DG loading requirements 
(kW and kVAR) specified in SRs 3.8.1.3 and 3.8.1.13. These loading 
requirements have no impact on or relationship to the probability of 
any of the initiating events assumed for the accidents previously 
evaluated. Therefore, the proposed change does not involve a 
significant increase in the probability of any accident previously 
evaluated. Furthermore, since the proposed loading requirements 
bound the maximum expected loading for the DGs, SRs 3.8.1.3 and 
3.8.1.13 will continue to demonstrate that the DGs are capable of 
performing their safety function. Since the proposed change does not 
adversely affect the capability of the DGs to perform their safety 
function, the outcomes of the accidents previously evaluated (i.e., 
radiological consequences) will not be affected. Therefore, the 
proposed change does not involve a significant increase in the 
consequences of any accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    No. The proposed change affects only the DG loading requirements 
(kW and kVAR) specified in SRs 3.8.1.3 and 3.8.1.13. The proposed 
change will not introduce any new equipment or create new failure 
modes for existing equipment. Other than the reduced loading 
requirements for the DGs, the proposed change will not affect or 
otherwise alter plant operation. The DGs will remain capable of 
performing their safety function. No other safety related or 
important to safety equipment will be affected by the proposed 
change. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed change reduces the loading requirements of SRs 
3.8.1.3 and 3.8.1.13. With one exception, the new loading 
requirements are consistent with the latest regulatory guidance 
found in Regulatory Guide (RG) 1.9, Revision 3, ``Selection, Design, 
and Qualification of Diesel-Generator Units Used as Standby (Onsite) 
Electric Power Systems at Nuclear Power Plants,'' July 1993. The one 
exception to RG 1.9, the loading requirements for the 2-hour portion 
of the endurance and margin test (SR 3.8.1.13), will require testing 
at loads in excess of 105 percent of the maximum expected load as 
opposed to 105 percent of the continuous duty rating. Testing for at 
least 2 hours at 105 percent of the maximum expected load will 
continue to demonstrate adequate margin, and it will reduce wear and 
tear on the DGs due to testing. Reduction in wear and tear should 
inherently increase the reliability of the DGs. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia.

[[Page 43781]]

    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of application for amendments: June 24, 1999 (TS 99-05)
    Brief description of amendments: The proposed amendments would 
change the Sequoyah Units 1 and 2 Technical Specification (TS) 
requirements by clarifying and changing the surveillance requirements 
for the ice weight in the ice condenser baskets. This request is a 
lead-plant change for all Westinghouse-designed ice condenser plants 
and will be incorporated into the Improved Standard Technical 
Specifications (ISTSs), if approved.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority, the licensee, has provided its analysis of the issue of no 
significant hazards consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed TS amendments discussed below cannot increase the 
probability of occurrence of any analyzed accident because they are 
not the result or cause of any physical modification to the ice 
condenser structures, and for the current design of the ice 
condenser, there is no correlation between any credible failure and 
the initiation of any previously analyzed accident.
    Regarding the consequences of analyzed accidents, the proposed 
amendment provides for consistency with the ISTSs by: (1) requiring 
the actions if one or more ice condenser ice baskets are determined 
to weigh below the minimum specified value to be made a part of the 
TS surveillance requirement (SR) instead of being located in the 
bases, and (2) relocating the ice basket selection methodology into 
the bases. This ensures consistent interpretation of the 
requirements of the TS in accordance with the ISTSs. The 
clarification of the response required if one or more ice baskets in 
a given bay are determined to be underweight ensures sufficient ice 
is maintained in each bay to prevent early meltout in a local zone 
following a design basis accident (DBA) and that the required 
overall ice weight is maintained in the ice condenser. The 
relocation of the ice basket selection methodology to the bases does 
not result in any change to the intent or implementation of this 
portion of the TSs since plant procedures ensure the requirements of 
the bases of the TSs are correctly implemented. Additionally, the 
clarification that the weight requirement is applicable to the 
beginning of the cycle does not change the present intent of the TS, 
but ensures there is no confusion, since the weight at the end of 
the operating cycle may be less than that specified in the SR due to 
sublimation. This does not result in a change to the intent or 
implementation of the TS since a sublimation allowance was provided 
in the original SR weight requirement. These clarifications do not 
result in any [effect] on plant equipment or operation and the 
actions taken during the implementation of the revised TS will be 
the same as prior to the revision. Therefore, the clarification of 
these requirements will not increase the consequences of any 
accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The inclusion of the action required for an underweight ice 
basket in the TS SR, instead of in the bases of the TS, provides for 
the consistent interpretation of the requirement. The clarification 
of the response required if one or more ice baskets in a given bay 
are determined to be underweight ensures sufficient ice is 
maintained in each bay to prevent early meltout in a local zone 
following a DBA and that the required overall ice weight is 
maintained in the ice condenser. The relocation of the ice basket 
selection methodology to the bases does not result in any change to 
the intent or implementation of this portion of the TSs since plant 
procedures ensure the requirements of the bases of the TSs are 
correctly implemented. Additionally, the clarification that the 
weight requirement is applicable to the beginning of the cycle does 
not change the present intent of the TS, but ensures there is no 
confusion, since the weight at the end of the operating cycle may be 
less than that specified in the SR due to sublimation. This does not 
result in a change to the intent or implementation of the TS since a 
sublimation allowance was provided in the original SR weight 
requirement. The operation, design and maintenance of the ice 
condenser and its associated equipment will not change as a result 
of these clarifications. Therefore, the implementation of these 
clarifications will not create the possibility of accidents or 
equipment malfunctions of a new or different kind from any 
previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed amendment allows for the consistent interpretation 
of the required actions if an ice basket is determined to weigh less 
than the required minimum. The inclusion of these actions in the TS 
SR instead of in the TS bases assures the correct actions will be 
taken as intended by the TSs. The clarification of the response 
required if one or more ice baskets in a given bay are determined to 
be underweight ensures sufficient ice is maintained in each bay to 
prevent early meltout in a local zone following a DBA and that the 
required overall ice weight is maintained in the ice condenser. The 
relocation of the ice basket selection methodology to the bases does 
not result in any change to the intent or implementation of this 
portion of the TSs since plant procedures ensure the requirements of 
the bases of the TSs are correctly implemented. Additionally, the 
clarification that the weight requirement is applicable to the 
beginning of the cycle does not change the present intent of the TS, 
but ensures there is no confusion, since the weight at the end of 
the operating cycle may be less than that specified in the SR due to 
sublimation. This does not result in a change to the intent or 
implementation of the TS since a sublimation allowance was provided 
in the original SR weight requirement. The proposed clarifications 
do not result in or have any [effect] on the operation, design, or 
maintenance of any plant equipment. Thus the design limits for the 
continued safe function of the containment structure following a DBA 
are not exceeded due to this change; therefore, the proposed 
amendment does not involve a reduction in a margin of safety.

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

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

    Date of amendment request: June 25, 1999 (TS 99-004).
    Description of amendment request: The proposed amendment would 
revise the Watts Bar Nuclear Plant Unit 1 Technical Specifications (TS) 
and associated TS Bases for Limiting Condition for Operation (LCO) 
3.7.1, Main Steam Safety Valves, to provide a new requirement to reduce 
the Power Range Neutron Flux-High reactor trip setpoints when two or 
more main steam safety valves (MSSVs) per steam generator are 
inoperable. This proposal is based on a generic change developed by the 
Westinghouse Owners Group (WOG), TSTF-235, Revision 1, which has been 
approved by the NRC staff.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 43782]]


    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change to TS LCO 3.7.1 requires a reduction of the 
Power Range Neutron Flux-High reactor trip setpoints to a 
corresponding power level depending on the number of inoperable 
MSSVs. The change is based on and consistent with an industry 
sponsored change (TSTF-235, Revision 1) which has been reviewed and 
accepted by the NRC staff.
    Although plant procedures currently require resetting the high 
flux trip, it is not a TS requirement. The proposed amendment will 
provide a more appropriate barrier to prevent the plant from being 
operated under a non-conservative technical specification action 
statement in a region where multiple inoperable MSSVs coincident 
with a reactivity insertion event such as an inadvertent rod cluster 
control assembly (RCCA) bank withdrawal could result in 
overpressurization of the secondary system.
    No change is made in the probability of initiating accident, 
i.e., RCCA bank withdrawal, and by requiring the reactor trip 
setpoint reduction, a potential mismatch between core power and 
turbine load without sufficient steam relief capacity is eliminated. 
Therefore, the change requested by this amendment actually decreases 
the consequences of an accident previously evaluated (without credit 
for procedure actions to reduce the trip setpoints).
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Without crediting existing plant procedures, the addition of the 
proposed TS change prevents the plant from being operated in a 
region where an overpressurization of the main steam system is 
postulated to potentially occur. The proposed change assures that 
the existing FSAR [Final Safety Analysis Evaluation Report] accident 
analysis remains bounding for events that challenge the relieving 
capacity of the MSSVs. Since the addition of the TS action adds a 
more appropriate administrative barrier to prevent operation in an 
undesired region and because the change is bounded by the current 
accident analysis described in the FSAR, a new or different kind of 
accident has not been created as a result of this license amendment.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed TS change eliminates a non-conservative TS action 
to prevent the plant from being operated in a region where an 
overpressurization of the main steam system is postulated to 
potentially occur. Since the addition of the TS action adds a more 
effective administrative barrier to prevent operation in an 
undesired region and because the change is bounded by the existing 
FSAR accident analysis, the margin of safety has actually increased 
for the proposed change. For these reasons, the proposed amendment 
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.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
    NRC Section Chief: Sheri R. Peterson.

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

    Date of amendment request: July 8, 1999.
    Description of amendment request: The amendment request proposes to 
increase the allowable values for the engineered safety features 
actuation system (ESFAS) loss-of-power 4 kV undervoltage trips in the 
current Technical Specifications (TSs) Table 3.3-4 (functional units 
8.a and 8.b) and in Surveillance Requirement (SR) 3.3.5.3 of the 
improved TSs. The word ``nominal'' is also being added to describe the 
trip setpoint in SR 3.3.5.3 and in the Bases of the improved TSs. The 
improved TSs were issued in Amendment 123 dated March 31, 1999, but 
have not yet been implemented.
    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 staff has reviewed the licensee's analysis against 
the standards of 10 CFR 50.92(c). The NRC staff's review is presented 
below.
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The reactor protection system performance will remain within the 
bounds of the previously performed accident analysis. The protection 
systems will continue to function in a manner consistent with the plant 
design basis. The proposed changes will not affect any of the analysis 
assumptions for any of the accidents previously evaluated. The proposed 
changes will not affect the probability of any event initiators nor 
will the proposed changes affect the ability of any safety related 
equipment to perform its intended function. There is no change to the 
technical specification trip setpoints; therefore, there is no 
degradation in the performance of nor an increase in the number of 
challenges imposed on safety related equipment assumed to function 
during an accident situation and be no change to normal plant operating 
parameters or accident mitigation capabilities. The allowable values 
and the trip setpoints in the protection system proposed to be changed 
are not initiators of accidents previously evaluated.
    Based on the above evaluation, these 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.
    There are no changes in the method by which any safety related 
plant system performs its safety function. The normal manner of plant 
operation remains unchanged because the methodology to determine the 
allowable value and the trip setpoints remains unchanged. The increase 
in allowable value for the trip setpoints still provides margin between 
the nominal trip setpoint and allowable value while taking into account 
worst case 4.16 kV Class 1E system (NB) bus voltages that could be 
possible during steady state loss-of-coolant accident (LOCA) 
conditions. The change in allowable value for the undervoltage 
protection functions does not impact the systems capability to:
    a. Trip the 4.16 kV preferred normal and alternate bus feeder 
breakers to remove the deficient power source to protect the Class 1E 
equipment from damage;
    b. Shed all loads from the bus except the Class 1E 480 Vac load 
centers and centrifugal charging pumps to prepare the buses for re-
energization by the load shedder and emergency load sequencer (LSELS); 
and
    c. Generate a emergency diesel generator (EDG) start signal.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result of 
the proposed changes. The allowable values and the trip setpoints in 
the protection system proposed to be changed are not initiators of 
accidents. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The undervoltage protection functions are to:

[[Page 43783]]

    a. Trip the 4.16 kV preferred normal and alternate bus feeder 
breakers to remove the deficient power source to protect the Class 1E 
equipment from damage;
    b. Shed all loads from the bus except the Class 1E 480 Vac load 
centers and centrifugal charging pumps to prepare the buses for re-
energization by the load shedder and emergency load sequencer (LSELS); 
and
    c. Generate a EDG start signal.
    The proposed changes do not affect the acceptance criteria for any 
analyzed event nor is there a change in the safety analysis limit. 
There will be no effect on the manner in which safety limits or 
engineered safety features actuation system settings are determined nor 
will there be any affect on those plant systems necessary to assure the 
accomplishment of the above protection functions. Therefore, there will 
not be a significant reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Section Chief: Stephen Dembek.

Previously Published Notice of Consideration of Issuance of 
Amendment to Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, And Opportunity for a Hearing

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

    Date of amendment request: June 30, 1999.
    Brief description of amendment: The proposed amendment would revise 
Technical Specification (TS) 3/4.7.5 of the current TSs by adding a 
temporary action statement that would allow the plant to operate for up 
to 12 hours with an inlet temperature up to but less than 95 deg.F.
    Date of individual notice in Federal Register: July 15, 1999 (64 FR 
38221).
    Expiration date of individual notice: August 16, 1999.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendment: February 26, 1999
    Brief description of amendment: This amendment changes the Table 
Notations for Technical Specification (TS) Table 3.3-4, ``Engineered 
Safety Features Actuation System Instrumentation Trip Setpoints.'' 
Specifically, the time constants used in the lead-lag controller for 
Steam Line Pressure--Low (Table item 1.e) and in the rate-lag 
controller for Negative Steam Line Pressure Rate--High (Table item 4.e) 
have been revised.
    Date of issuance: July 28, 1999.
    Effective date: July 28, 1999.
    Amendment No.: 89.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 24, 1999 (64 FR 
14280).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 28, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

    Date of application of amendments: October 2, 1998, as supplemented 
November 20, 1998, December 21, 1998, and May 13, 1999.
    Brief description of amendments: The amendments revised the Updated 
Final Safety Analysis Report related to an unreviewed safety question 
regarding the use of a small amount of containment overpressure to 
ensure sufficient net positive suction head for the reactor building 
spray and low pressure injection pumps during the post loss of coolant 
accident recirculation phase.
    Date of Issuance: July 19, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--305; Unit 2--305; Unit 3--305.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: June 16, 1999 (64 FR 
32288).
    The November 20, 1998, December 21, 1998, and May 13, 1999, letters 
provided clarifying information that did not change the scope of the 
October 2, 1998, application and the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 19, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501

[[Page 43784]]

West South Broad Street, Walhalla, South Carolina.

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

    Date of application for amendment: May 28, 1998.
    Brief description of amendment: Changes the Crystal River Unit 3 
(CR-3) licensing bases to incorporate Generic Letter 87-11, 
``Relaxation in Arbitrary Intermediate Pipe Rupture Requirements,'' and 
NUREG/CR-2913, ``Two-Phase Jet Loads,'' as part of the licensing basis 
for CR-3.
    Date of issuance: July 27, 1999.
    Effective date: As of the date of issuance, to be incorporated into 
the Final Safety Analysis Report at the time of its next update.
    Amendment No.: 181.
    Facility Operating License No. DPR-72: Amendment approves changes 
to the Final Safety Analysis Report.
    Date of initial notice in Federal Register: July 15, 1998 (63 FR 
38200).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 27, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.

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

    Date of amendment request: March 1, 1999, as supplemented by 
letters dated March 10, 1999, June 8, 1999, and June 23, 1999.
    Brief description of amendment: The amendment changes the Cooper 
Nuclear Station Technical Specifications to revise the calibration 
frequency of the reactor recirculation flow transmitters from once 
every 184 days to once every 18 months.
    Date of issuance: July 26, 1999.
    Effective date: July 26, 1999, to be implemented within 30 days.
    Amendment No.: 179.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 7, 1999 (64 FR 
17027) The March 10, June 8, and June 23, 1999, letters provided 
additional clarifying information and updated TS pages. This 
information was within the scope of the original Federal Register 
notice and did not change the staff's initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Auburn Memorial Library, 1810 
Courthouse Avenue, Auburn, NE 68305.

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

    Date of amendment request: September 28, 1998, as supplemented by 
letter dated March 12, 1999.
    Brief description of amendment: The amendment authorizes the 
revision to the licensing basis as described in the Updated Safety 
Analysis Report (USAR) to incorporate the modification for overriding 
the containment isolation actuation signal to the reactor coolant 
system letdown flow containment isolation valves.
    Date of issuance: July 22, 1999.
    Effective date: July 22, 1999, and shall be implemented in the next 
periodic update to the USAR in accordance with 10 CFR 50.71(e).
    Amendment No.: 191.
    Facility Operating License No. DPR-40. The amendment revised the 
Updated Safety Analysis Report.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 64119) The March 12, 1999, supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noticed, and did not change the staff's initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 22, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.

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

    Date of amendment request: January 29, 1999.
    Brief description of amendment: The amendment revises Technical 
Specifications 2.10.2(1) and 2.10.2(3) and deletes Figure 2-11 to 
relocate three cycle specific parameters to the Core Operating Limits 
Report.
    Date of issuance: July 27, 1999.
    Effective date: July 27, 1999.
    Amendment No.: 192.
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1999 (64 
FR 9193) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 27, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.

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: January 25, 1999.
    Brief description of amendments: Revise technical specifications 
surveillance requirement frequencies for the emergency diesel generator 
maintenance inspection outages, the 24-hour endurance run and the hot 
restart test from 18 to 24 months.
    Date of issuance: July 29, 1999.
    Effective date: Both units, as of date of issuance and shall be 
implemented within 30 days.
    Amendment Nos.: 136 and 101.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1999 (64 
FR 9196).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 29, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.

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

    Date of application for amendments: February 2, 1999, as 
supplemented on April 26, 1999.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 5.6, ``Fuel Storage, Criticality,'' to change the 
maximum unirradiated fuel assembly enrichment value for new fuel 
storage from 4.5 to 5.0 weight percent Uranium-235 and to allow the use 
of equivalent criticality control to that provided by the current TS 
requirement of 2.35 milligrams of Boron-10 per linear inch loading in 
the Integral Fuel Burnable Absorber pins.
    Date of issuance: July 21, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.

[[Page 43785]]

    Amendment Nos.: 223 and 204.
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 10, 1999 (64 FR 
11965).
    The April 26, 1999, letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 21, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.
    Dated at Rockville, Maryland, this 4th day of August 1999.
    For the Nuclear Regulatory Commission.
Suzanne C. Black,
Deputy Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 99-20545 Filed 8-10-99; 8:45 am]
BILLING CODE 7590-01-P