[Federal Register Volume 64, Number 144 (Wednesday, July 28, 1999)]
[Notices]
[Pages 40903-40914]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-19133]


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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 Pub. L. 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

[[Page 40904]]

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 3, 1999, through July 16, 1999. The 
last biweekly notice was published on July 14, 1999 (64 FR 38022).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By August 27, 1999, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating 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 much 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.

[[Page 40905]]

    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendment request: March 26, 1999.
    Description of amendment request: The proposed change provides a 
Required Action and Completion Time for the Ultimate Heat Sink (UHS) in 
the event that service water temperature exceeds the current 95 deg.F 
surveillance limit. It involves an allowance to continue operation for 
a period of 8 hours with the UHS at a temperature greater than the 
temperature limits provided in Technical Specification (TS) Limiting 
Condition of Operation 3.7.8, ``Ultimate Heat Sink (UHS)'' and provides 
an upper UHS temperature limit beyond which plant shutdown is required.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Carolina Power & Light (CP&L) Company has evaluated the proposed 
Technical Specification change and has concluded that it does not 
involve a significant hazards consideration. The conclusion is in 
accordance with the criteria set forth in 10 CFR 50.92. The bases 
for the conclusion that the proposed change does not involve a 
significant hazards consideration are discussed below.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change does not involve any physical alteration of 
plant systems, structures or components. The proposed change will 
allow plant operation for a short period of time when the service 
water temperature exceeds 95 deg.F. If the service water temperature 
is restored within the allowed time, a plant shutdown is not 
required. This minimizes plant transients, which reduces the 
probability of a reactor trip and the resulting challenges to 
mitigating systems. A service water temperature of up to 99 deg.F 
does not increase the failure rate of systems, structures or 
components because the systems, structures, and components are 
designed for higher temperatures than at which they operate.
    The Service Water (SW) System temperature is not assumed to be 
an initiating condition of any accident evaluated in the safety 
analysis report. Therefore, the allowance of a limited time for 
service water temperature to be in excess of 95 deg.F does not 
involve an increase in the probability of an accident previously 
evaluated in the safety analysis report (SAR). The SW System 
supports operability of safety related systems used to mitigate the 
consequences of an accident. The service water temperature is not 
expected to increase significantly beyond 95 deg.F due to the 
limited time allowed by the proposed change in conjunction with the 
generally slow rate of temperature increase experienced from thermal 
changes in Lake Robinson. The capability of components to perform 
their safety related function is not affected up to a service water 
temperature of 99 deg.F with the exception of the Containment Air 
Recirculation Fan Coolers. The heat removal capacity of the 
Containment Air Recirculation Fan Coolers is not expected to be 
significantly reduced by a small increase in service water 
temperature. If heat removal is not significantly reduced, 
containment pressure and leakage will not be significantly 
increased, and the doses from containment leakage will not be 
significantly increased. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated in the SAR.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve any physical alteration of 
plant systems, structures or components. A service water temperature 
of up to 99 deg.F does not introduce new failure mechanisms of 
systems, structures or components not already considered in the SAR 
because the systems, structures, and components are designed for 
higher temperatures than at which they operate. Therefore, the 
possibility of a new or different kind of accident from any accident 
previously evaluated is not created.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change will allow a small increase in service water 
temperature above the design basis limit for the SW System and delay 
by 8 hours the requirement to shutdown the plant when the service 
water system design limit is exceeded. There are design margins 
associated with systems, structures and components that are cooled 
by the service water system that are affected. The capability of 
components to perform their safety related function is not affected 
up to a service water temperature 99 deg.F with the exception of the 
Containment Air Recirculation Fan Coolers. The Containment Air 
Recirculation Fan Coolers remove heat from containment to mitigate 
containment pressure and temperature following a MSLB (main 
streamline break) inside containment or a Large Break LOCA (loss-of-
coolant accident) inside containment. An increase in service water 
temperature in excess of the design limit due to hot weather 
conditions is expected to be small due to the limited time allowed 
by the proposed change in conjunction with the generally slow rate 
of temperature increase experienced from thermal changes in Lake 
Robinson. Therefore, the effect on the Containment Air Recirculation 
Fan Coolers' heat removal capacity and the resulting containment 
pressure and temperature is expected to be small. Therefore, there 
is no significant reduction in margin of safety associated with this 
change.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Sheri R. Peterson.

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

    Date of amendment request: June 29, 1999.
    Description of amendment request: This amendment request proposes 
to increase the notch testing surveillance interval of partially 
withdrawn control rods in Technical Specification Surveillance 
Requirement 3/4.3.C,

[[Page 40906]]

``Reactivity Control--Control Rod Operability,'' from an interval of 
once in 7 days to once in 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:

    Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    The proposed change extends the Surveillance Frequency for 
partially withdrawn control rods. The change does not affect 
equipment design or operation. The affected Surveillance is not 
considered to be an accident initiator. Therefore, this change will 
not significantly increase the probability of an accident previously 
evaluated. Furthermore, extension of the Surveillance Frequency will 
not impact the ability to perform its function following an 
accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The extension of the Surveillance Frequency does not involve 
physical modification to the plant and does not introduce a new mode 
of operation.
    Therefore, the change does 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 change in the Surveillance Frequency only provides a minor 
reduction in the probability of finding an inoperable control rod. 
Most of the control rods will continue to be tested on the current 
Frequency. However, if one stuck rod is identified, all rods must be 
checked promptly.
    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 proposed to determine that 
the requested amendments involve no significant hazards 
consideration.

    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: April 1, 1999.
    Description of amendment request: The proposed license amendment 
would modify the Technical Specifications (TSs) to incorporate certain 
improvements from the Revised Standard Technical Specifications for B&W 
Plants (NUREG-1430) that would add limiting conditions for operation 
action statements, make surveillance requirements more consistent with 
the revised standard TSs, correct conflicts or inconsistencies from 
earlier TS revisions, correct administrative errors, and revise the 
spent fuel pool sampling from monthly and after adding chemicals to 
weekly.
    The staff's proposed no significant hazards determination below 
does not address the licensee's proposed changes with respect to a high 
pressure injection system operation in a low temperature overpressure 
environment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated. The proposed amendment makes administrative corrections, 
adds conditions to the limiting conditions of operation [LCOs], 
revises selected time clocks and surveillance requirements 
consistent with NUREG 1430, and adds a time clock to a unique LCO. 
These changes have no effect on the plant design or operation. The 
reliability of systems and components relied upon to prevent or 
mitigate the consequences of accidents previously evaluated is not 
degraded by proposed changes. Therefore, operation in accordance 
with the proposed amendment does not involve a significant increase 
in the probability of occurrence 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, because no new 
accident initiators would be created.
    3. Operation of the facility in accordance with the proposed 
amendment will not involve a significant reduction in a margin of 
safety because no changes to plant operating limits or limiting 
safety system settings are proposed.

    The NRC staff has reviewed the licensee's analysis and based on the 
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: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 22037.
    NRC Section Chief: S. Singh Bajwa.

Power Authority of The State of New York, Docket No. 50-286, Indian 
Point Nuclear Generating Unit No. 3, Westchester County, New York

    Date of amendment request: June 4, 1999.
    Description of amendment request: This application for amendment to 
the Indian Point 3 Technical Specifications (TSS) proposes to revise 
the definition of operating personnel in section 6.2.2.g to make it 
consistent with the Standard Technical Specifications and to remove a 
footnote.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licenses has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No, these TS changes are administrative in nature. Removing the 
statement in section 6.2.2.g that defines on shift operating 
personnel and adding a new paragraph consistent with the Standard 
Technical Specifications is an administrative line item change that 
follows NRC guidance. The current statement is not needed because TS 
Table 6.2.1 defines the minimum operations shift crew composition 
and commitments to Table B-1 of NUREG-0654 defines the minimum 
staffing requirements for each function area.
    The change to TS 6.2.2.i is administrative in nature. The 
statement that reads, ``For the period ending three years after 
restart from the 1993/1994 Performance Improvement Outage, the 
Operations Manager will be permitted to have held a SRO [senior 
reactor operator] license at a Pressurized Water Reactor other than 
Indian Point Unit 3'', was a relaxation of the requirements of 
6.2.2i.
    Therefore, these changes will not increase the probability or 
consequences of an accident previously evaluated, because they are 
administrative and affect neither accident initiation or mitigation.
    2. Does the proposed license amendment create the possibility of 
a new or different

[[Page 40907]]

kind of accident from any accident previously evaluated?
    No, these TS changes are administrative in nature. Removing the 
statement in section 6.2.2.g that defines on shift operating 
personnel and adding a new paragraph consistent with the Standard 
Technical Specifications is an administrative line item change that 
follows NRC guidance. The current statement is not needed because TS 
Table 6.2-1 defines the minimum operations shift crew composition 
and commitments to Table B-1 of NUREG-0654 defines the minimum 
staffing requirements for each function area.
    The change to TS 6.2.2.i is administrative in nature. The 
statement that reads, ``For the period ending three years after 
restart from the 1993/1994 Performance Improvement Outage, the 
Operations Manager will be permitted to have held a SRO license at a 
Pressurized Water Reactor other than Indian Point Unit 3'', was a 
relaxation of the requirements of 6.2.2.i.
    These changes are administrative, and do not affect how the 
plant is operated. They also follow the guidance of the Standard 
Technical Specifications. Therefore, these changes will 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?
    No, these TS change is administrative in nature. Removing the 
statement in section 6.2.2.g that defines on shift operating 
personnel and adding a new paragraph consistent with the Standard 
Technical Specification is an administrative line item change that 
follows NRC guidance. The current statement is not needed because TS 
Table 6.2-1 defines the minimum operations shift new composition and 
commitments to Table B-1 of NUREG-0654 defines the minimum staffing 
requirements for each function area.
    The change to TS 6.2.2.i is administrative in nature. The 
statement that reads, ``For the period ending three years after 
restart from the 1993/1994 Performance Improvement Outage, the 
Operations Manager will be permitted to have held a SRO license at a 
Pressurized Water Reactor other than Indian Point Unit 3'', was a 
relaxation of the requirements of 6.2.2.i.
    These changes are administrative, and do not affect how the 
plant is operated. They also follow the guidance of the Standard 
Technical Specifications. Therefore, these 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 proposed to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for Licensee; Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    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

    Description of amendment requests: The licensee proposed changes to 
Technical Specification (TS) 3.3.5 ``ESFAS Instrumentation'' to include 
restrictions on operation with a channel of the refueling water storage 
tank level-low input to the recirculation actuation signal (RAS) and 
the steam generator pressure-low input or steam generator pressure 
difference-high input to the emergency feedwater actuation signal 
(EFAS) in the tripped condition. The current TS allows plant operation 
in this condition indefinitely. The licensee has determined that 
unacceptable consequences could result from a spurious trip of RAS or 
EFAS due to operation with a channel in trip condition. The licensee 
states that the proposed TS changes would improve plant operational 
safety and, thereby, reduce plant risk.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    This change provides limits for operating with a channel of the 
Refueling Water Storage Tank (RWST) Level-Low input in the 
Recirculation Actuation Signal (RAS) or the Steam Generator (SG) 
Pressure-Low or SG Pressure Difference (SGPD)-High input to the 
Emergency Feedwater Actuation Signal (EFAS) in trip.
    As a result of this change, the potential for an inadvertent 
actuation of either of these two signals is reduced. The proposed 
Completion Times are based on Probabilistic Risk Assessment (PRA) 
considerations, and are conservative compared to the current 
unlimited Completion Times.
    The consequences of an inadvertent actuation of EFAS or RAS are 
unaffected by this change.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    (2) Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    This proposed change provides additional time limits on 
operation with a channel of the RWST Level-Low input to RAS or the 
SG Pressure-Lower SGPD-High inputs to EFAS in trip. Operation in 
this condition is currently allowed indefinitely. The proposed 
restrictions reduce the possibility of an inadvertent actuation of 
RAS or EFAS, and do not allow operation in any configuration not 
currently allowed by the Technical Specifications (TSs).
    Therefore, this proposed change will not create the possibility 
of a new or different kind of accident from any accident that has 
been previously evaluated.
    (3) Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed change provides additional time limits on operation 
with a channel of the RWST Level-Low input to RAS or the SG 
Pressure-Low or SGPD-High inputs to RAS or EFAS in trip. The 
proposed limits are conservative compared to the current 
requirements, where the time limit is unrestricted. The overall 
impact of the change will be [an] increase in the margin of safety.
    Therefore, there will be no significant reduction in a margin of 
safety as a result of this change.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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.

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant, 
Unit 2, Hamilton County, Tennessee

    Date of application for amendments: June 7, 1999 (TS 99-09).
    Brief description of amendments: The proposed amendment would 
change the Sequoyah Unit 2 Technical Specification (TS) requirements by 
adding a new temporary Figure 3.4-1a and temporary footnotes to TS 
3.4.8, ``Specific Activity,'' Table 4.4-4, and to corresponding Bases 
in order to raise the reactor coolant specific activity limit to 1.0 
microcurie per milligram Dose Equivalent iodine-131 for the remainder 
of Unit 2 Cycle 10 operation.
    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

[[Page 40908]]

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 change increases the allowed reactor coolant 
specific activity for iodine-131 and decreases the leakage quantity 
that would be postulated to occur at the faulted steam generator 
(SG) during a main steam line break (MSLB) accident. The described 
changes will return these parameters to the same values under which 
the plant operated prior to the implementation of TS Change 98-02 
submitted on June 26, 1998. The June 26, 1998 submittal was a 
voluntary change that allowed for a greater leakage quantity during 
an MSLB accident as described in Generic Letter 95-05. Returning 
these parameters to their previous values does not affect or 
increase the probability of any accidents previously evaluated.
    An increase in the consequences of an accident would not occur 
because the proportional increase in reactor coolant specific 
activity, while proportionally decreasing the allowable primary-to-
secondary leakage during a postulated MSLB accident to values under 
which the plant was previously operated, was evaluated in [Topical 
Report No.] WCAP-13990 during the establishment of the original 
primary-to-secondary leak limits. No changes to the physical plant, 
to the plant operation, or maintenance practices have been 
implemented that would invalidate the limits defined in WCAP-13990.
    The control room dose, the low population zone dose, and the 
dose at the exclusion area boundary remain bounded by the acceptance 
criteria of the Updated Final Safety Analysis Report. Therefore, the 
proposed TS change does not result in an increase in the 
consequences of an accident previously analyzed.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS change does not alter the configuration of the 
plant. The changes do not directly affect plant operation. The 
change will not result in the installation of any new equipment or 
systems or the modification of any existing equipment or systems. No 
new operating procedures, conditions, or modes will be created by 
this proposed change. SG tube structural integrity, as defined in 
draft Regulatory Guide 1.121, remains unchanged. Therefore, the 
possibility of a new or different kind of accident from any accident 
previously evaluated is not created.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    Raising the allowed reactor coolant specific activity, while 
decreasing the allowed primary-to-secondary leakage during a 
postulated MSLB accident, keeps the amount of activity released to 
the environment unchanged. Design basis and offsite dose calculation 
assumptions remain satisfied. Therefore, the proposed change does 
not result in a significant reduction in the 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 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Sheri R. Peterson.

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

    Date of amendment request: May 24, 1999, as supplemented by letter 
dated July 9, 1999.
    Brief description of amendments: The proposed license amendments 
would remove several cycle-specific parameter limits from the Technical 
Specifications (TSs). These parameter limits would be added to the Core 
Operating Limits Report (COLR). Appropriate references to the COLR 
would be inserted in the affected TSs. In addition, the core safety 
limit curves would be replaced with safety limits more directly 
applicable to the fuel and fuel cladding fission product barriers. The 
affected Technical Specifications are: (1) TS 2.0, ``Safety Limits 
(SLs),'' (2) TS 3.3.1, ``Reactor Trip System Instrumentation 
Setpoints,'' (3) TS 3.4.1, ``RCS pressure temperature and flow from 
Nucleate Boiling (DNB) Limits,'' and (4) TS 5.6.5, ``Core Operating 
Limits Report.'' The May 24, 1999, application was previously noticed 
and published in the Federal Register on June 30, 1999 (64 FR 53213).
    The July 9, 1999, supplement provided proposed additional 
information that would: (a) Add the Reactor Core Safety Limit figures 
to the COLR, (b) clarify that the overpower N-16 setpoint remains in 
the TSs, and (c) reflect NRC approval of the topical reports used to 
determine the core operating limits presented in the COLR. The 
supplemental information is being noticed herein to address the issue 
of no significant hazards consideration.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes remove cycle-specific parameter limits from 
the Technical Specifications, add them to the list of limits 
contained in the Core Operating Limits Report (COLR), and revise the 
Administrative Controls section of the Technical Specifications. The 
proposed changes also insert the original minimum RCS flow limits 
into the Technical Specifications. The changes do not, by 
themselves, alter any of the parameter limits. The changes are 
administrative in nature and have no adverse effect on the 
probability of an accident or on the consequences of an accident 
previously evaluated. The removal of parameter limits from the 
Technical Specifications does not eliminate the requirement to 
comply with the parameter limits.
    The parameter limits in the COLR may be revised without prior 
NRC approval. However, [Technical] Specification 5.6.5c continues to 
ensure that the parameter limits are developed using NRC-approved 
methodologies and that applicable limits of the safety analyses are 
met. While future changes to the COLR parameter limits could result 
in event consequences which are either slightly less or slightly 
more severe than the consequences for the same event using the 
present parameter limits, the differences would not be significant 
and would be bounded by the requirement of specification 5.6.5c to 
meet the applicable limits of the safety analysis.
    Based on the above, addition of the minimum RCS flow limit into 
the Technical Specifications, removal of the parameter limits the 
Technical Specifications and the addition of the described limits in 
the COLR, thus allowing revision of the parameter limits without 
prior NRC approval, has no significant effect on the probability or 
consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes add the minimum RCS flow limit into the 
Technical Specifications, remove certain parameter limits from the 
Technical Specifications and add these limits to the list of limits 
in the COLR, thus removing the requirements for prior NRC approval 
of revisions to those parameters. The changes do not add new 
hardware or change plant operations and therefore cannot initiate an 
event nor cause an analyzed event to progress differently. Thus, the 
possibility of a new or different kind of accident is not created.
    3. Do the proposed changes involved a significant reduction in a 
margin of safety?
    The margin of safety is the difference between the acceptance 
criteria and the associated failure values. The proposed changes do 
not affect the failure values for any parameter. Though the accident 
analyses, all applicable limits (i.e., relevant event acceptance 
criteria as described in the NRC-approved analysis methodologies) 
are shown to be satisfied; therefore, there is no impact

[[Page 40909]]

on event acceptance criteria. Because neither the failure values nor 
the acceptance criteria are affected, the proposed change has no 
effect on 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: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: May 26, 1999.
    Description of amendment request: The licensee proposed revising 
the suppression pool water temperature surveillance requirements to 
specify monitoring the temperature every 5 minutes when performing 
testing that adds to the suppression pool. In addition, the licensee 
proposed revising the requirement to check the suppression chamber 
water level and temperature from ``once per shift'' to ``daily'' and 
specify that it is the average temperature that is checked.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided the NCR its analysis of the issue of no significant hazard 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment, will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    Vermont Yankee has determined that the proposed change will not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated. The proposed change revises the 
surveillance frequency for ``once per shift'' suppression pool water 
level and temperature monitoring. Additionally, the surveillance 
requirement for suppression pool water temperature monitoring when 
there are indications of relief valve operation that add heat to the 
suppression pool is also revised. The proposed change will revise the 
surveillance wording such that routine suppression pool monitoring will 
be ``daily'' and an operator will verify pool temperature every 5 
minutes only during testing that adds heat to the suppression pool. 
Also clarified, is that the parameter being monitored is ``average'' 
suppression pool water temperature.
    The consequence of an accident previously evaluated is not 
significantly increased since the initial suppression pool water 
temperature limit, which is an input valve for accident analyses, is 
not changed.
    The proposed change affects only surveillance requirements and does 
not require any hardware or equipment modification. Equipment 
operation, plant limiting conditions for operation, and accident 
analyses will be unchanged. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
accidents.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment, will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Vermont Yankee has determined that the proposed change does not 
create the possibility of a new or different kind of accident from any 
accident previously evaluated. The proposed change involves revision of 
Technical Specification surveillance requirements. There are no 
hardware modifications or equipment changes involved and operation of 
plant equipment will be unchanged. Thus, no new or different accident 
precursors will be created by this change.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment, will not involve a significant 
reduction in a margin or safety. VY has determined that the proposed 
change does not involve a significant reduction in a margin of safety. 
The proposed change involves revision of Technical Specification 
surveillance requirements. There are no hardware modifications or 
equipment changes involved and plant operation and accident analyses 
are unchanged. The initial suppression pool water temperature limit, 
which is an input value for accident analyses, is not changed. 
Therefore, the proposed change will not involve a significant reduction 
in the margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attoney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: June 29, 1999.
    Description of amendment request: The licensee proposed revising 
the leak rate requirements of Technical Specifications 3.7.A.4 and 
4.7.A.4 for the main steam line isolation valves. Specifically, a total 
leakage rate allowable value for the sum of the four main steam lines 
is proposed that is equal to four times the current individual main 
steam line isolation valve leakage rate allowable value. The individual 
main steam line isolation valve leakage rate allowable value is 
proposed to be one half of the total leakage rate allowable value.
    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 operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any of the parameters or conditions that contribute to the 
initiation of any accidents previously evaluated. Thus, the proposed 
change cannot increase the probability of any accident previously 
evaluated.
    The proposed change does not affect the leak-tight integrity of 
the containment structure that is designed to mitigate the 
consequences of a loss-of-coolant accident (LOCA). The primary 
containment must maintain functional integrity during and following 
the peak transient pressures and temperatures that result from any 
LOCA, thereby limiting fission product leakage following the 
accident. Because the proposed change does not alter any of the 
fission product lead rate assumptions used in the design basis LOCA 
analysis, the analyzed consequences of the Loss of Coolant Accident 
are not changed.
    The control room radiological habitability analysis uses as an 
input assumption main steam line leakage rate at four times the 
current Technical Specifications limit. An allowable value for total 
main steam line

[[Page 40910]]

leakage rate equivalent to four times the current Technical 
Specifications limit for a single main steam line isolation valve is 
being added by this change. Thus, there is no effect on the main 
control room radiological habitability calculation.
    Based on the above VY [Vermont Yankee] has concluded that the 
proposed change will not result in a significant increase in the 
probability or consequences of any accident previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any parameters or conditions that could contribute to the 
initiation of any accident. The methods of performing the tests are 
not changed. No new accident modes are created. No safety-related 
equipment or safety functions are altered as a result of this 
change. Restating the acceptance criteria while maintaining the 
assumptions of all affected calculations has no influence over nor 
does it contribute to, the possibility of a new or different kind of 
accident or malfunction from those previously evaluated.
    Based on the above VY has concluded that the proposed change 
will not create the possibility of a new or different kind of 
accident from those previously evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    Restating the acceptance criteria for the main steam line 
isolation valve leakage rate while maintaining the assumptions of 
all affected calculations does not impact the margin of safety. The 
0.6La maximum and minimum pathway leakage rate acceptance 
criteria provide the previously analyzed margin of safety. The 
testing method for determining the leak-tightness of the main steam 
line isolation valves has not changed. The leak rate test results 
are presently added to the Types B and C tests summation. The 
0.6La maximum and minimum pathway leak rate acceptance 
criteria and the proposed Technical Specifications requirements 
provide assurance that component degradation does not impact the 
assumptions used to determine, nor provide a reduction in, and the 
analyzed margin of safety.
    Based on the above VY has concluded that the proposed change 
will not cause 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: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trobridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: July 12, 1999.
    Description of amendment request: The amendment would revise the 
value for the Safety Limit Minimum Critical Power Ratio (SLMCPR) and 
delete the wording specifying these as Cycle 20 values.
    Basis for proposed no significant hazards consideration 
determination: As required by to CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment, will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The basis of the SLMCPR is to ensure no mechanistic fuel damages 
is calculated to occur if the limit is not violated. The new SLMCPR 
values preserve the existing margin to transition boiling and 
probability of fuel damage is not increased. The derivation of the 
revised SLMCPR for Vermont Yankee for incorporation into the 
Technical Specifications, and its use to determine plant and cycle-
specific thermal limits, have been performed using NRC approved 
methods. These plant-specific calculations are performing each 
operating cycle and if necessary, will require future changes to 
these values based upon revised core designs. The revised SLMCPR 
values do not change the method of operating the plant and have no 
effect on the probability of an accident initiating event or 
transient.
    Based on the above, Vermont Yankee has concluded that the 
proposed change will not result in a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment, will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed changes result only from a specific analysis for 
the Vermont Yankee core reload design and deletion of a cycle 
specific reference for the values. These changes do not involve any 
new or different method for operating the facility and do not 
involve any facility modifications. No new initiating events or 
transients result from these changes.
    Based on the above, Vermont Yankee has concluded that the 
proposed change will not create the possibility of a new or 
different kind of accident from those previously evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment, will not involve a 
significant reduction in a margin of safety.
    The new SLMCPR is calculated using NRC approved methods with 
plant and cycle specific parameters for the current core design. The 
SLMCPR value remains high enough to ensure that greater than 99.9% 
of all fuel rods in the core will avoid transition boiling if the 
limit is not violated, thereby preserving the fuel cladding 
integrity. The operating MCPR limit is set appropriately above the 
safety limit value to ensure margin when the cycle specific 
transients are evaluated.
    As a result, Vermont Yankee has determined that the proposed 
change will not result in a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

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

    Date of amendment request: June 22, 1999 (TSCR 210).
    Description of amendment request: The proposed amendments reflect 
changes to the Point Beach Nuclear Plant (PBNP) Units 1 and 2 Technical 
Specifications (TSs) in order to incorporate the Westinghouse 422V+ 
fuel assemblies into the PBNP reactor cores. Basis for proposed no 
significant hazards consideration determination: As required by 10 CFR 
50.91(a), the licensee has provided its analysis of the issue of no 
significant hazards consideration which is presented below:

    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not result in a significant increase in 
the probability or consequences of an accident previously evaluated.
    The accidents which are potentially affected by the parameters 
and assumptions associated with this amendment have been evaluated/
analyzed and all design standards and applicable safety criteria are 
met. The consideration of these changes does not result in a 
situation where the design and construction standards that were 
applicable prior to the change are altered. Therefore, the changed 
occurring with this amendment will not result in any additional 
challenges to plant equipment that could increase the probability of 
any previously evaluated accident.
    The proposed changes associated with this amendment do not 
affect plant systems such that their function in the control of

[[Page 40911]]

radiological consequences is adversely affected. The safety 
evaluation (included in Attachment 2 of this submittal) documents 
that the design standards and applicable safety criteria limits 
continue to be met and therefore fission barrier integrity is not 
challenged. The proposed changes have been shown not to adversely 
affect the response of the plant to postulated accident scenarios. 
Existing system and component redundancy and operation is not being 
changed by these proposed changes. These changes will therefore not 
affect the mitigation of the radiological consequences of any 
accident described in the FSAR [final safety analysis report].
    In some cases, the results of the revised radiological analyses 
are greater than those of the current FSAR analysis. In other cases, 
the new and old analyses are not directly comparable because the 
radiological bases for the new analyses have been upgraded to meet 
more current NRC requirements. However, in all cases, the calculated 
doses are well within the regulatory acceptance criteria and do not 
constitute an unacceptable significant increase in consequences. 
Since the actual plant configuration, performance of systems, and 
initiating event mechanisms are not being changed as a result of 
this evaluation, the probability or consequences of an accident 
previously evaluated is not significantly increased.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The possibility for a new or different type of accident from any 
accident previously evaluated is not created as a result of this 
amendment. The changes described in the amendment are supported by 
the analyses and evaluations described in Attachment 2 (safety 
evaluation). The evaluation of the effects of the proposed changes 
indicate that all design standards and applicable safety criteria 
limits are met. These changes therefore do not cause the initiation 
of any new or different accident nor create any new failure 
mechanisms.
    All equipment important to safety will continue to operate as 
designed. Component integrity is not challenged. The changes do not 
result in any event previously deemed incredible being made 
credible. The changes do not result in more adverse conditions or 
result in any increase in the challenges to safety systems. 
Therefore, operation of the Point Beach Nuclear Plant in accordance 
with the proposed amendments will not create the possibility of a 
new or different type of accident from any accident previously 
evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not involve a significant reduction in 
a margin of safety.
    The proposed changes do not involve a significant reduction in 
the margin of safety. Existing component redundancy is not being 
changed by these proposed changes. There are no new or significant 
changes to the initial conditions contributing to accident severity 
or consequences. The margin of safety is maintained by assuring 
compliance with acceptance limits reviewed and approved by the NRC. 
Since all of the appropriate acceptance criteria for the various 
analyses and evaluations have been met as discussed in Attachment 2 
(Safety Evaluation) of this submittal and provided for information 
in Attachment 4 (PBNP FSAR Chapter 14 ``Safety Analysis'' changes 
required as a result of the analyses performed for the upgraded 
fuel) of this submittal, by definition there has not been a 
significant reduction of any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: July 1, 1999 (TSCR 214).
    Description of amendment request: The proposed amendments reflect a 
change to Point Beach Nuclear Plant (PBNP) Units 1 and 2 Technical 
Specification (TS) Section 15.5.4. The amendment request proposes to 
remove one of the two separate methods for verifying the acceptability 
of reactor fuel for placement and storage in the spent fuel pool.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes are administrative only in that they remove 
the ability to use the reference Koo method for 
determining the acceptability of fuel for placement and storage in 
the spent fuel pool and new fuel storage vault at the Point Beach 
Nuclear Plant. Use of the remaining approved method and requirements 
ensure that fuel placed or stored in the spent fuel pool and new 
fuel storage vault continues to be in accordance with their 
respective design and licensing basis. That is, fuel in the storage 
array will continue to meet the design basis requirement that 
Keff remain less than 0.95. No modifications are being 
made to the spent fuel pool and its cooling system or to the new or 
spent fuel storage racks. Since the design basis of the fuel and 
storage racks continue to be met, operation in accordance with the 
proposed amendments cannot create a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    No physical modifications are being made to the spent fuel pool 
and cooling system or to the new or spent fuel storage racks. All 
design basis requirements for ensuring the safe storage of fuel in 
the spent fuel pool continue to be met. Therefore, operation in 
accordance with the proposed amendments cannot create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not create a significant reduction in a 
margin of safety.
    Technical Specification requirements for placing and storing 
fuel in the spent fuel pool continue to ensure that the design basis 
requirement, Keff for the fuel array in the spent fuel 
pool and new fuel storage remains less than 0.95, is maintained. The 
existing margin of safety established by this design requirement is 
maintained. Therefore, operation in accordance with the proposed 
amendments cannot create 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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.
    Attorney for license: John H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

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

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

    Date of amendment request: May 27, 1999, as supplemented by letter 
dated May 28, 1999

[[Page 40912]]

    Description of amendment request: The proposed amendments would add 
a footnote to Technical Specification (TS) 4.8.2.1e, ``D.C. Sources-
Operating,'' which would, on a one-time basis for Unit 1 Battery 
BT1ED2, allow the licensee to substitute a performance discharge test 
``*  *  * in lieu of the battery service test required by Specification 
4.8.2.1d, twice within a 60 month interval.''
    Date of publication of individual notice in Federal Register: June 
14, 1999. (64 FR 31881).
    Expiration date of individual notice: July 14, 1999.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas.

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 1, 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.

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

    Date of application for amendment: December 21, 1998, as 
supplemented on January 28, February 18, April 2, April 15, and April 
16, 1999.
    Brief description of amendment: This amendment makes changes to 
Facility Operating License No. DPR-35, the Technical Specifications, 
and Materials License No. 20-07626-04 to reflect the transfer of the 
licenses from Boston Edison Company to Entergy Nuclear Generation 
Company.
    Date of issuance: July 13, 1999.
    Effective date; As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 181.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: January 26, 1999 (64 FR 
3984). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 29, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

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

    Date of application for amendment: March 3, 1999.
    Brief description of amendment: The amendment modified Technical 
Specification Table 4.6-3, ``Reactor Vessel Material Surveillance 
Program Withdrawal Schedule.'' The amendment changed the withdrawal 
schedule for the upcoming reactor vessel surveillance capsule pull from 
approximately 15 effective full power years to approximately 18 
effective full power years.
    Date of issuance: July 15, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 182.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 19, 1999 (64 FR 
27316). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 15, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

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

    Date of application for amendments: December 1, 1997, supplemented 
August 26, 1998.
    Brief description of amendments: Revised the Technical 
Specifications (TS), Appendix B, Environmental Protection Plan (Non-
Radiological), to implement the terms and conditions of the incidental 
Take Statement included in the Biological Opinion issued by the 
National Marine Fisheries Service, regarding endangered sea turtles.
    Date of Issuance: July 2, 1999.
    Effective Date: July 2, 1999.
    Amendment Nos.: 162 and 103.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the TS.
    Date of initial notice in Federal Register: December 31, 1997 (62 
FR 68305). The supplemental letter dated August 26, 1998, provided 
clarifying information that did not change the original no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 2, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

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

    Date of application for amendment: June 5, 1998, as supplemented 
January 13, 1999.
    Brief description of amendment: The proposed revision to the 
Millstone Unit 3 licensing basis would address a recent steam generator 
tube rupture (SGTR) analysis that was determined to be an unreviewed 
safety question. The SGTR analyses described in the Final Safety 
Analysis Report (FSAR) include an offside dose analysis and a margin to 
overfill analysis. Both of the analyses have been updated. The offsite 
dose analysis was updated to reflect a larger capacity for the steam 
generator atmospheric dump valve (ADV) and a decrease in the operator 
response time to close the ADV block valve. The

[[Page 40913]]

margin to overfill analysis was updated to reflect a new single 
failure.
    Date of issuance: July 2, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 172.
    Facility Operating License Nos. DPR-49: Amendments authorizes 
revisions to the FSAR,
    Date of initial notice in Federal Register: July 1, 1998 (63 FR 
35992).
    The January 13, 1999, supplemental letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 2, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut

Northeast Nuclear Energy Company, et al., Docket No. 50-335, Millstone 
Nuclear Power Station, Unit No. 2, New Location County, Connecticut

    Date of application for amendment: March 19, 1999.
    Brief description of amendment: The amendment relocated Technical 
Specifications Sections 3.3.3.2, ``Instrumentation, Incore Detectors,'' 
3.3.3.3, ``Instrumentation, Seismic Instrumentation,'' and 3.3.3.4, 
``Instrumentation, Meteorological Instrumentation,'' to the Millstone, 
Unit No. 2, Technical Requirements Manual. Index page V and TS Bases 
have been revised to reflect the above relocations.
    Dated of issuance: July 13, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 237.
    Facility Operating License Nos. DPR-65: Amendment Revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 21, 1998 (64 FR 
19560).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 13, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

PP&L, Inc., Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric 
Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of application for amendment: November 26, 1999, which was 
superseded by letter dated June 1, 1998, as supplemented by letters 
dated October 30, 1998, March 29, 1999, April 20, 1999, and May 28, 
1999.
    Brief description of amendment: These amendment would replace the 
current ultimate heat sink average water temperature limit for all 
combination of plant operations.
    Dated of issuance: July 6, 1999.
    Effective date: Both units, effective as of date of issuance and 
shall be implemented within 30 days.
    Amendment Nos.: 182 and 156.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 20, 1998 (63 FR 
27764). The October 30, 1998, March 29, 1999, April 20, 1999, and May 
28, 1999, letters 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 6, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilker-Barre, PA 18701.

PP&L, Inc., Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric 
Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of application for amendment: November 23, 1999.
    Brief description of amendments: The amendments modified the 
Susquehanna Steam Electric Station, Units 1 and 2, Technical 
Specifications limiting condition for operation, 3.8.3, and 
surveillance requirements, 3.8.3.1, to increase the minimum fuel oil 
storage tank volume ranges.
    Dated of issuance: July 7, 1999.
    Effective date: Units 1 and 2, as of date of issuance and shall be 
implemented within 30 days.
    Amendment Nos.: 183 and 157.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 27, 1999 (64 FR 
4160).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 7, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Osterhout Free Library, 
Reference, Department, 71 South Franklin Street, Wikes-Barre, PA 18701.

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 application for amendments: October 17, 1997, as 
supplemented March 2 and November 28, 1998.
    Brief description of amendments: These amendments authorize changes 
to the updated Final Safety Analysis Report (FSAR) to permit 
installation of digital radiation monitors for both the containment 
purge isolation and the control room isolation signals.
    Date of issuance: July 12, 1999.
    Effective date: July 12, 1999; implementation shall include 
submission by the licensee of the revised description authorized by 
these amendments with the next update of the FSAR in accordance with 10 
CFR 50.71(e).
    Amendment Nos.: Unit 2-154; Unit 3-145.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the FSAR.
    Date of initial notice in Federal Register: January 28, 1998 (63 FR 
4324).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 12, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P.O. Box 19557, Irvine, California 92713.

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

    Date of application for amendments: January 12,1999, as 
supplemented by letters dated May 11, and June 30, 1999.
    Brief description of amendments: The amendment revised Technical 
Specification 3/4.7.5, Ultimate Heat Sink, by adding a new action 
statement to be used in the event the plant inlet water temperature 
exceeds 90 deg. F. The amendment is effective only through September 
30, 1999, and is only for the current TSs. The amendment is also 
limited to a maximum plant inlet water temperature of 94 deg. F. The 
proposal to raise this temperature to 95 deg. F will be addressed in a 
future letter.

[[Page 40914]]

    Date of issuance: July 8, 1999.
    Effective date: July 8, 1999, shall be implemented within 30 days 
of the date of issuance.
    Amendment No.: 125.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1999 ( 64 
FR 9203). The May 11 and June 30, 1999, supplemental letters provided 
additional clarifying information, did not expand the scope of the 
application as originally noticed and did not change the staff's 
original proposed no significant hazards consideration determination, 
except that the licensee proposed a maximum plant inlet water 
temperature of 95 deg. F. where the letters of January and May 11, 
1999, proposed only 94 deg. F. The amendment is limited to a maximum 
temperature of 94 deg. F.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 8, 1999.
    No significant hazards consideration comments received: No.
    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.

    Dated at Rockville, Maryland, this 21st day of July 1999.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management Office of Nuclear 
Reactor Regulation.
[FR Doc. 99-19133 Filed 7-27-99; 8:45 am]
BILLING CODE 7590-01-M