[Federal Register Volume 64, Number 56 (Wednesday, March 24, 1999)]
[Notices]
[Pages 14278-14292]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-7032]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 1, 1999, through March 12, 1999. The 
last biweekly notice was published on March 10, 1999 (64 FR 11958).

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 April 23, 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

[[Page 14279]]

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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of amendments request: December 16, 1998.
    Description of amendments request: The proposed amendment would 
revise Technical Specification (TS) 3.8.1, ``AC Sources--Operating,'' 
and TS 3.3.7, ``Diesel Generator (DG)--Loss of Voltage Start (LOVS).'' 
The proposed amendment will (1) change Condition G of TS 3.8.1 to 
ensure that the appropriate actions will be taken to prevent double 
sequencing of safety-related loads, and (2) change TS 3.3.7 to ensure 
that the setpoint allowable values for the degraded voltage and the 
loss of voltage relays reflect the required function of the relays. 
Basis for proposed no significant hazards consideration determination: 
As required by 10 CFR 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment will change Condition G of Technical 
Specification 3.8.1. These changes will ensure that the appropriate 
actions will be taken to prevent double sequencing of safety-related 
loads. This change is required to assure the capability of the 
offsite circuits ``to effect a safe shutdown and to mitigate the 
effects of an accident'' in accordance with Regulatory Guide 1.93. 
The proposed amendment will also change the setpoint allowable 
values for the degraded voltage and the loss of voltage relays in 
Technical Specification Surveillance Requirement (SR) 3.3.7.3. The 
proposed changes do not involve any physical changes to plant 
equipment. The actions required by the TS amendment will identify 
when an offsite circuit does not meet its required capability and 
provides actions to restore the required capability. The proposed 
changes are intended to identify and correct the conditions (voltage 
and/or loading) required to prevent the possibility of a double 
sequencing event. Therefore, this change ensures that power will be 
supplied to the ESF [engineered safety feature] loads following a 
loss of offsite power event described in UFSAR [Updated Final Safety 
Analysis Report] 15.2.6.1. For other events discussed in the UFSAR, 
the electrical distribution system is an event mitigator. This 
change will ensure that the electrical distribution system will 
continue to meet this requirement. The proposed changes will not 
effect the function of the DG loss of voltage start as required by 
the design basis and safety analysis. Therefore, the proposed change 
does not involve a significant increase in the probability of an 
accident previously evaluated.
    The proposed changes do not involve any physical changes to 
plant equipment. The proposed changes ensure that appropriate 
controls are in place to prevent a double sequencing event. The 
proposed changes consider the factors in preventing a double 
sequencing event such as pretrip voltage, load, number of units on 
line, and number of transmission lines in service. These are factors 
which could affect post trip voltage. The actions associated with 
this change will identify and mitigate the condition where an 
offsite circuit does not meet its required capability and, as such, 
do not result in new or revised accident sequences. The proposed 
changes will not effect the function of the DG loss of voltage start 
as required by the design basis and safety analysis. Therefore, the 
proposed change does not involve a significant increase in the 
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.
    The proposed amendment will change Condition G of Technical 
Specification 3.8.1. These changes will ensure that the appropriate 
actions will be taken to prevent double sequencing of safety-related 
loads. This change is required to assure the capability of the 
offsite circuits ``to effect a safe shutdown and to mitigate the 
effects of an accident'' in accordance with Regulatory

[[Page 14280]]

Guide 1.93. The proposed amendment will also clarify the setpoint 
allowable values for the degraded voltage and the loss of voltage 
relays in Technical Specification Surveillance Requirement (SR) 
3.3.7.3. The proposed changes do not change the operation of any 
system or equipment, nor do they create a new type of malfunction. 
The proposed changes prevent double sequencing and do not create the 
possibility of any other malfunction. The actions associated with 
this change will identify and mitigate the condition where an 
offsite circuit does not meet its required capability. The proposed 
changes will not effect the function of the DG loss of voltage start 
as required by the design basis and safety analysis. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed amendment will change Condition G of Technical 
Specification 3.8.1. These changes will ensure that the appropriate 
actions will be taken to prevent double sequencing of safety-related 
loads. This change is required to assure the capability of the 
offsite circuits ``to effect a safe shutdown and to mitigate the 
effects of an accident'' in accordance with Regulatory Guide 1.93. 
The proposed amendment will also change the setpoint allowable 
values for the degraded voltage and the loss of voltage relays in 
Technical Specification Surveillance Requirement (SR) 3.3.7.3. The 
proposed changes ensure that the units will be in conformance with 
GDC 17, Electric Power Systems (basis for TS 3.8.1). The required 
actions of the proposed change will ensure that the single failure 
analyses and safety analysis are maintained. The actions associated 
with this change will identify and mitigate the condition where an 
offsite circuit does not meet its required capability. The proposed 
changes ensure that the bases for the current TS are maintained. The 
proposed changes will not effect the function of the DG loss of 
voltage start as required by the design basis and safety analysis. 
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 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Project Director: William H. Bateman.

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: February 26, 1999.
    Description of amendment request: The proposed amendment would 
revise 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.) are 
t1 greater than or equal to 50 seconds and t2 
greater than or equal to 5 seconds. The proposed amendment would revise 
t2 to less than or equal to 5 seconds. Also, the time 
constant used in the rate-lag controller for Negative Steam Line 
Pressure Rate--High (Table item 4.e.) is less than or equal to 50 
seconds. The proposed amendment would revise this time constant to 
greater than or equal to 50 seconds.
    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.
    Correcting the time constants will ensure conservative 
calibration of the Engineered Safety Feature Actuation System 
instrumentation. 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. 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.
    Correcting the time constants will ensure conservative 
calibration of the Engineered Safety Feature Actuation System 
instrumentation. 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 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.
    Correcting the time constants will ensure conservative 
calibration of the Engineered Safety Feature Actuation System 
instrumentation. 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 Project Director: Cecil Thomas.

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

    Date of application for amendment: March 26, 1997.
    Brief description of amendment: The amendment revises modifies 
Technical Specification sections 3.6 and 4.5 by removing the list of 
containment isolation valves in accordance with Generic Letter 91-08, 
``Removal of Components Lists from Technical Specifications,'' dated 
May 6, 1991, and by revising requirements related to containment 
pressure and containment temperature. Additionally, several editorial 
changes are made to emulate the format and content of NUREG-1432, 
``Standard Technical Specifications, Combustion Engineering Plants.''
    Date of issuance: February 22, 1999.
    Effective date: February 22, 1999.
    Amendment No.: 184.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 17, 1997 (62 
FR 66136)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 22, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.

[[Page 14281]]

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

    Date of amendment request: September 3, 1997.
    Description of amendment request: The proposed amendment includes 
the following changes to the station technical specification (TS):
    (a) TS Action Statement 3.14a is replaced by a revised condition 
description for TS Action Statement 3.17.1.6 in the instrumentation 
systems section. Also, the maximum control room temperature at which a 
shutdown must be initiated is revised from 120  deg.F [degrees 
Fahrenheit] to 90  deg.F, and a time limit for reaching the hot 
shutdown condition is specified;
    (b) TS 3.14b is replaced with two limiting conditions for operation 
(LCOs), 3.14.1 and 3.14.2, addressing, respectively, the filtration and 
cooling functions of the CRHVAC [control room heating, ventilation, and 
air conditioning] system. These proposed LCOs emulate the standard TS 
(NUREG 1432) for control room ventilation;
    (c) TS Table 4.2.3 surveillance requirement (SR) number 3, 
verification of control room temperature, is moved to SR Table 4.17.1, 
for the reactor protection system (RPS); and
    (d) other administrative changes.
    The licensee classified each change as either administrative or 
more restrictive. An administrative change is editorial in nature, 
involves only movement of requirements within the TS without affecting 
their technical content, or clarifies existing TS requirements. A more 
restrictive change adds new requirements, or revises existing 
requirements resulting in more conservative or additional operational 
restrictions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes to TS 3.14a and TS 3.14b constitute either 
new, or more restrictive requirements that provide additional assurance 
that equipment conforms to the plant design basis and will operate 
reliably when called upon. These changes represent additional 
restrictions on plant operation that enhance safety and are consistent 
with the standard TS. The proposed change to TS Table 4.2.3 of moving 
SR item number 3 to TS Table 4.17.1, and other administrative changes 
are editorial in nature or involve the reorganization or reformatting 
of TS requirements without affecting technical content or operational 
restrictions. The proposed changes do not result in any substantive 
change in operating requirements or the intent of these requirements, 
and are consistent with the Commission's regulations. Therefore, these 
changes cannot involve a significant increase in 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 previously evaluated?
    The proposed changes to TS 3.14a and TS 3.14b constitute either 
new, or more restrictive requirements that provide additional assurance 
that equipment conforms to the plant design basis and will operate 
reliably when called upon. These changes represent additional 
restrictions on plant operation that enhance safety and are consistent 
with the standard TS. The proposed change to TS Table 4.2.3 of moving 
SR number 3 to TS Table 4.17.1, and other administrative changes are 
editorial in nature or involve the reorganization or reformatting of TS 
requirements without affecting technical content or operational 
restrictions. The proposed changes do not result in any substantive 
change in operating requirements or the intent of these requirements, 
and are consistent with the Commission's regulations. Therefore, these 
changes cannot create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The proposed changes to TS 3.14a and TS 3.14b constitute either 
new, or more restrictive requirements that provide additional assurance 
that equipment conforms to the plant design basis and will operate 
reliably when called upon. These changes represent additional 
restrictions on plant operation that enhance safety and are consistent 
with the standard TS. The proposed change to TS Table 4.2.3 of moving 
SR number 3 to TS Table 4.17.1, and other administrative changes are 
editorial in nature or involve the reorganization or reformatting of TS 
requirements without affecting technical content or operational 
restrictions. The proposed changes do not result in any substantive 
change in operating requirements or the intent of these requirements, 
and are consistent with the Commission's regulations. Therefore, these 
changes cannot involve 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 location: Van Wylen Library, Hope 
College, Holland, Michigan 49423-3698.
    Attorney for licensee: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Project Director: Cynthia A. Carpenter.

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 amendment request: March 1, 1999.
    Description of amendment request: The proposed amendments would 
revise Oconee Nuclear Station, Units 1, 2, and 3 Improved Technical 
Specification (ITS) 3.3.8 to only require two channels for the reactor 
coolant system hot leg temperature function. The current TSs require 
two channels per loop. This requirement was incorrectly specified 
during the ITS conversion.
    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 change modifies ITS Table 3.3.8-1 to only require 
two channels for RCS [Reactor Coolant System] Hot Leg Temperature 
Function. These instruments provide indication only and are not 
considered as initiators of any analyzed event. The proposed change 
does not involve a physical alteration of the plant. No new or 
different equipment is being installed, and no installed equipment 
is being operated in a new or different manner. No set points for 
parameters which initiate protective or mitigative action are being 
changed. Therefore, the proposed changes do 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 kind of accident previously evaluated:
    The proposed change does not involve a physical alteration of 
the plant. No new or different equipment is being installed, and no

[[Page 14282]]

installed equipment is being operated in a new or different manner. 
No set points for parameters which initiate protective or mitigative 
action are being changed. As a result, no new failure modes are 
being introduced. Therefore, this proposed amendment will not create 
the possibility of any new or different kind of accident.
    3. Involve a significant reduction in a margin of safety.
    The margin of safety for PAM [post accident monitoring] 
instrumentation is based on the availability and capability of the 
instrumentation to provide the required operator information. The 
proposed change maintains requirements within the safety analyses 
and licensing basis and has no effect on the availability and 
capability of the PAM function. Therefore, the 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: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.
    Attorney for licensee: Ann W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC.
    NRC Project Director: Herbert N. Berkow.

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 amendment request: March 1, 1999.
    Description of amendment request: The proposed amendments to 
Improved Technical Specification (ITS) 3.9, ``Refueling Operations,'' 
Subsection 3.9.3, ``Containment Penetrations,'' Limiting Condition for 
Operation 3.9.3.b would add a Note to state that the emergency air lock 
door is not required to be closed when it is sealed with a temporary 
cover plate. The temporary cover plate contains penetrations that are 
used for such refueling outage services as cables, pneumatic tubing, 
and hoses.
    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:

    This proposed change has been evaluated against the standards in 
10 CFR 50.92 and has been determined to involve no significant 
hazards, in 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?
    The proposed change allows the use of a temporary cover plate as 
a seal for the emergency air lock during refueling operations in 
lieu of an air lock door. Duke [Duke Energy Corporation] analyses 
for Oconee Nuclear Station (ONS) does not credit containment 
closure. Therefore, use of the temporary cover plate does not affect 
offsite doses, which were previously calculated to be well within 10 
CFR 100 limits. As such, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The fuel handling accident inside containment analyses discussed 
in the Updated Final Safety Analysis Report section 15.11 bound the 
proposed change. No new or different type of accident will occur 
because of the temporary cover plate placement.
    3. Involve a significant reduction in a margin of safety.
    Placing the temporary cover plate in the emergency air lock will 
still meet the intent of containment closure. The building pressure 
does not increase during a fuel handling accident and fission 
products will be contained. The fuel handling accident inside 
containment analyses does not credit containment closure for 
reducing offsite dose. As such, 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: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: February 24, 1999.
    Description of amendment request: The proposed amendments would 
change Technical Specification (TS)
3/4.7.4 to remove the restriction to monitor the Ultimate Heat Sink 
(UHS) temperature only in the Intake Cooling Water (ICW) bay and prior 
to the ICW pumps. This change would permit the option of monitoring the 
UHS temperature after the ICW pumps but prior to the component cooling 
water heat exchangers, which is considered to be equivalent to 
temperature monitoring before the ICW pumps.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. 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.
    The method of monitoring the Ultimate Heat Sink temperature is 
not considered in, and has no effect on, the probability of any type 
of accident initiating sequence. The proposed changes will permit 
other means of monitoring the Ultimate Heat Sink that have been 
evaluated to be equivalent to the current method permitted. As the 
monitoring will continue to be performed by equal means, the 
consequences of any accident previously evaluated will not be 
affected.
    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 accident previously evaluated.
    The proposed change will permit other means of monitoring the 
Ultimate Heat Sink temperature, which will be equal to the methods 
currently employed. The continued monitoring of this variable by 
equivalent means cannot create the possibility of a new or different 
type of accident.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The Ultimate Heat Sink temperature is an input assumption used 
in the accident analysis and in evaluation of component design. This 
temperature limit is not being altered by this change, only the 
permissible means of monitoring this variable. As any new methods 
employed are expected to be equivalent to those currently used, no 
reduction in any margin of safety will result.

    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: Florida International 
University, University Park, Miami, Florida 33199.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Project Director: Cecil O. Thomas.

[[Page 14283]]

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

    Date of amendment request: February 2, 1999.
    Description of amendment request: The proposed amendment revises 
the Technical Specifications (TS) to expand the scope of systems and 
test requirements considered under TS 4.5.4 ``Engineered Safeguards 
Feature (ESF) Systems Leakage,'' and increases the maximum allowable 
leakage for those portions of the ESF system outside containment. The 
proposed amendment also includes revised the Bases for TS 3.15.3, 
``Auxiliary and Fuel Handling Building Air Treatment System,'' to 
clarify system design requirements and accident analysis 
considerations.
    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 or consequences of an accident previously evaluated. No 
physical modifications which would change structures, systems or 
components are proposed by this TSCR [technical specification change 
request] for surveillance changes in Technical Specification 4.5.4 
and its Bases. The proposed increase in the ESF Systems leakage rate 
acceptance limit has no effect on the performance of ESF systems 
during a DBA [design basis accident]. The proposed changes are 
supported by a revised MHA [maximum hypothetical accident] dose 
calculation using updated X/Q values and calculation assumptions. 
The MHA dose consequence analysis yields dose results that are below 
the 10 CFR 100 guidelines for both the EAB [exclusion area boundary] 
and LPZ [low population zone]. The calculated Control Room 
Habitability Evaluation does not exceed the permissible annual 
occupational exposure limit of 50 Rem to the thyroid as specified in 
10 CFR 20.1201(a)(ii). In addition, the potential thyroid exposure 
can be mitigated by the availability of self-contained breathing 
apparatus and potassium iodide. Therefore, the changes would not 
involve a significant increase in the consequences of accidents 
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. This TSCR does not 
involve any physical modifications that would affect structures, 
systems, or components, nor does it involve any changes in plant 
operation.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. This TSCR does not involve changes to the Technical 
Specification defined Safety Limits, Limiting Conditions for 
Operation, and does not involve any change to safety system 
setpoints for operation. Therefore, the proposed changes do not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment 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 20037.
    NRC Project Director: Elinor G. Adensam.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: February 12, 1999.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TS) to (1) allow reactor vessel 
hydrostatic and leakage tests without maintaining primary containment 
integrity, (2) establish a limit and a surveillance requirement on 
reactor coolant activity when reactor coolant temperature is above 
212 deg.F, the reactor is not critical, and primary containment has not 
been established, and (3) correct a punctuation error.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes do not increase the probability of an 
accident since reactor vessel hydrostatic and leakage tests would be 
performed with the reactor vessel nearly water solid, at nominal 
operating pressure, not critical and at low decay heat values which 
minimizes the energy stored in the reactor vessel. Under this 
proposed change a limit on reactor coolant activity is established 
that provides adequate assurance that the consequences of a large 
primary system break during reactor vessel hydrostatic and leakage 
test conditions will be conservatively bounded by the consequences 
of a postulated main steam line break outside of primary 
containment. Low pressure emergency core cooling systems are 
required to be operable during reactor vessel hydrostatic and 
leakage test providing assurance that adequate core cooling can be 
achieved to preclude fuel failures and subsequent increases in 
reactor coolant activity in the event of a large primary system 
break. The reduced stored energy in the reactor vessel and proposed 
limit on reactor coolant activity ensures there is no increase in 
the probability or consequences of an accident previously evaluated.
    The proposed amendment will not create the possibility of a new 
or different kind of accident from any accident previously analyzed.
    The proposed changes do not introduce any new accident 
initiators or failure mechanisms since the changes do not involve 
any changes to the structures, systems, or components. They also do 
not involve any change to the operation of systems, and alter 
procedures only to the extent that 212 deg.F may be exceeded during 
reactor vessel hydrostatic and leakage testing without maintaining 
primary containment integrity. Without maintaining primary 
containment integrity, a large primary system break during a reactor 
vessel hydrostatic or leakage test would result in the same kind of 
accident as would a main steam line break outside primary 
containment during normal operation. Therefore, the proposed TS 
change does not create the possibility of a new or different kind of 
accident, from any accident previously evaluated.
    The proposed amendment will not involve a significant reduction 
in the margin of safety.
    Since reactor vessel hydrostatic and leakage tests are performed 
nearly water solid, at nominal operating pressure, not critical and 
at low decay heat values, the stored energy in the reactor vessel 
during testing will be low. Under these conditions, the potential 
for failed fuel and a subsequent increase in coolant activity is 
minimized. Therefore, the proposed Technical Specification 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. In addition, correction to the punctuation error is strictly 
a grammatical change and has no effect on the three standards of 10 CFR 
50.92(c). Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and

[[Page 14284]]

Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter.

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: February 8, 1999.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 4.5.3.2.b to allow the option of using 
closed and disabled automatic valves to provide the necessary isolation 
function when performing safety injection and charging pump testing in 
Modes 4, 5, and 6 (hot shutdown, cold shutdown, and refueling) for low 
temperature over pressurization protection.
    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.
    In Mode 4 with the RCS [reactor coolant system] coolant 
temperature less than 312  deg.F or in Modes 5 and 6 there is a 
potential risk of low temperature overpressurization. Mass additions 
of coolant by the safety injection and charging pumps could cause 
such an event to the extent that these pump flows exceed the ability 
of a single over pressure protection relief valve to protect the 
system. In order to eliminate this potentiality provisions are made 
to allow a maximum of one pump to be in service with the other pumps 
disabled except for testing. Further provisions are made to assure 
that a pump being tested can not inject into the vessel. The 
proposed change merely adds an alternate method of providing this 
assurance in addition to that currently provided by closing the 
manual discharge valves. The proposed change offers an equivalent 
means of affording the required protection.
    Based upon the above, the proposed change will not increase the 
probability or consequences of an accident previously analyzed.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed changes do not require any change in the operation 
of the plant. A minor configuration change is involved in that a[ ] 
disabled automatic valve in the flow path will be used in lieu of 
the manual valve to provide protection. Specifically, no new 
hardware is being added to the plant as part of the proposed change, 
no existing equipment is being modified, and no significant changes 
in operations are being introduced. Therefore, these changes will 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed change will not alter any assumptions, initial 
conditions, or results of any accident analyses. The proposed change 
maintains the level of protection. The change will, therefore, 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 Project Director: Elinor G. Adensam.

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

    Date of amendment request: March 1, 1999.
    Description of amendment request: The proposed amendment would 
revise the Ginna Station Improved Technical Specifications battery cell 
parameters limit for specific gravity Surveillance Requirement (SR) 
3.8.6.3 and SR 3.8.6.6.
    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 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated. The change is only 
to correct an error in the determination of the minimum limiting 
value for specific gravity of the station batteries. This does not 
increase the probability of an accident previously evaluated since 
the battery specific gravity is only a measure of the state of 
charge of the battery and the batteries themselves are not an 
accident initiator. The proposed minimum value for specific gravity, 
based on the NUREG-1431 guidance, gives a higher assurance that the 
battery has sufficient capacity. Therefore, the probability or 
consequences of an accident previously evaluated is not 
significantly increased.
    (2) Operation of Ginna Station in accordance with the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated. The proposed change 
does not involve a physical alteration of the plant (i.e. no new or 
different type of equipment will be added) or changes in the methods 
governing normal plant operation. The change only involves 
implementing a more conservative minimum limiting value for the 
battery cell parameter of specific gravity. 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 
change does not involve a significant reduction in a margin of 
safety. The proposed change only corrects an error in the 
determination of the limiting value for specific gravity. The error 
is being corrected by using a more conservative value as determined 
by the guidance of NUREG-1431. Therefore, this 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: 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 Project Director: S. Singh Bajwa.

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

    Date of amendment request: February 18, 1999.
    Description of amendment request: The proposed amendment would 
revise Virgil C. Summer Nuclear Station (VCSNS) Technical Specification 
(TS) 3/4.4.9 Reactor Coolant System Pressure/Temperature Limits to 
incorporate the new Pressure/Temperature (PT) Limits curves consistent 
with reactor vessel specimen analysis results. Additionally, the 
proposed amendment would revise the Pressure/Temperature Limits Bases 
section to accurately reflect current industry standards and 
regulations.
    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 proposed changes revise the Pressure/Temperature Limits 
Curves to provide curves that reflect the results of the analysis 
performed on reactor vessel surveillance

[[Page 14285]]

specimen W. This analysis was performed using NRC approved 
methodology as documented in WCAP 14040-NP-A, dated January, 1996. 
These curves provide the limits for operation of the Reactor Coolant 
System during heat up, cool down, criticality, and hydrotesting. The 
limits protect the reactor vessel from brittle fracture by 
separating the region of acceptable operation from the region where 
brittle fracture is postulated to occur. Failure of the reactor 
vessel is not a VCSNS design basis accident, and, in general, 
reactor vessel failure has a low probability of occurrence and is 
not considered in the safety analysis.
    Therefore, the change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes revise the Pressure/Temperature Limits 
Curves, Section 3/4.4.9, to incorporate the results of the analysis 
performed on reactor vessel specimen W. There are no plant design 
changes or significant changes in any operating procedures. This 
change adjusts the heat up and cool down curves to reflect the shift 
in nil-ductility reference temperature of the reactor vessel as a 
result of neutron embrittlement. Therefore, the change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does this change involve a significant reduction in margin of 
safety? The proposed changes revise the Pressure/Temperature Limits 
Curves, Section 3/4.4.9, to incorporate the results of the analysis 
performed on reactor vessel specimen W. The new PT curves ensure 
that the 10 CFR 50 Appendix G, requirements are not exceeded during 
normal operation including Reactor Coolant System transients during 
heat up, cool down, criticality, and hydrotesting. The new PT curves 
were prepared, using approved NRC methodology, for a projected 
reactor vessel neutron exposure of 32 EFPY [effective full power 
years].
    The new curves shift to more conservative operating limitations, 
thus providing increased margin against non-ductile fractures. Since 
administrative limits remain in place to ensure that 10 CFR 50 
Appendix G limits are not challenged, the margin of safety described 
in the TS Bases is not reduced by the proposed change. Therefore, 
the 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: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180.
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment requests: May 8, 1996, as supplemented by letter 
dated January 13, 1999.
    Description of amendment requests: The January 13, 1999, 
supplemental letter added an additional change to the technical 
specifications (TS) to incorporate an additional restriction to the 
time required to close containment when reactor coolant system (RCS) 
water level is reduced during a refueling outage. This additional 
restriction adds a limitation that containment must be able to be 
closed within the calculated time to boil, if it is less than the 
current four hour requirement. The January 13, 1999, letter supplements 
the staff's proposed no significant hazards consideration determination 
evaluation that was published on September 11, 1996 (61 FR 47978).
    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 licensee's analysis of the issue of no significant 
hazards consideration on the supplemental change is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Item 6 conservatively restricts the completion time to ensure 
containment closure is achieved prior to the water in the cavity 
boiling, in the event of a Loss of Shutdown Cooling. This 
restriction is already a self imposed requirement at San Onofre 
Units 2 and 3. Incorporating it in the Technical Specification only 
serves to highlight the importance of this requirement.
    This change captures all periods of time when the time to boil 
following a Loss of Shutdown Cooling is less than 4 hours. Having 
this requirement cannot initiate an accident. However, this 
requirement reduces the consequences of a Loss of Shutdown Cooling 
Accident when the time to boil is less than 4 hours.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Item 6 conservatively restricts the completion time to ensure 
containment closure is achieved prior to the water in the cavity 
boiling, in the event of a Loss of Shutdown Cooling. This 
restriction is already a self imposed requirement at San Onofre 
Units 2 and 3. Incorporating it in the Technical Specification only 
serves to highlight the importance of this requirement.
    This restriction cannot initiate an accident.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Item 6 conservatively restricts the completion time to ensure 
containment closure is achieved prior to the water in the cavity 
boiling, in the event of a Loss of Shutdown Cooling. This 
restriction is already a self imposed requirement at San Onofre 
Units 2 and 3. Incorporating it in the Technical Specification only 
serves to highlight the importance of this requirement.
    This change increases the margin of safety provided by the 
Technical Specification by specifying that the containment must be 
closed within 4 hours or within the calculated time to boil, 
whichever is less. This change revises the Technical Specification 
to specifically recognize the importance of ensuring containment 
closure is achieved prior to boiling in the reactor vessel, upon a 
loss of shutdown cooling.

    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, P.O. Box 19557, Irvine, California 92713.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, P. O. Box 800, Rosemead, California 91770.
    NRC Project Director: William H. Bateman.

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

    Date of amendment requests: December 22, 1998.
    Description of amendment requests: The proposed amendment would 
modify the technical specifications (TS) to add a reference to allow 
use of Westinghouse laser-welded steam generator (SG) tube sleeving. 
The proposed amendment also provides typographical and editorial 
corrections.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Steam generator tubes, tube plugging, and tube failures are 
considered in the analysis of

[[Page 14286]]

accidents in the Updated Final Safety Analysis Report (UFSAR). The 
steam generator tube rupture accident analysis considered the 
failure of a steam generator tube. Also, inadvertent opening of a 
steam generator dump valve (IOSGDV), loss of condenser vacuum 
(LOCV), loss of coolant accidents (LOCAs), and feed water line break 
(FWLB) accident analyses carry assumptions regarding steam generator 
tube plugging. In each case, the addition of steam generator tube 
sleeves to repair defective tubes will not change the probability or 
consequences of any accident previously evaluated.
    The sleeve configurations have been designed, analyzed, and 
tested in accordance with the American Society of Mechanical 
Engineers (ASME) code requirements, and mechanical testing has shown 
that the sleeve and sleeve joints provide margin above acceptance 
limits. Ultrasonic testing (UT) and eddy current testing (ECT) are 
used to verify the adequacy of welds. Tests have demonstrated that 
tube collapse will not occur due to postulated LOCA loadings.
    The probability or consequences of any accident previously 
evaluated is not increased because any leakage through the sleeve 
assembly is fully bounded by the existing steam generator tube 
rupture analysis included in the San Onofre Unit 2 and 3 Updated 
Final Safety Analysis Report. Additionally, any reactor coolant flow 
restriction from sleeving is addressed by a ratio of number of 
sleeved tubes to be equal to a plugged tube.
    Therefore, the proposed sleeving repair process will not involve 
an increase in the probability or consequences of any previously 
evaluated accident.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The sleeves are captured within the steam generator tubes by 
hard rolling and welding and as such are not able to physically 
affect other parts of the system. The failure of a sleeve is 
identical to the failure of the parent tube which has been 
previously analyzed.
    The use of a sleeve to span the area of degradation of the steam 
generator tube restores the structural and leakage integrity of the 
tubing to meet the original design requirements. Structural analysis 
of the sleeve assembly shows that the requirements of the ASME code 
are met. Mechanical testing has demonstrated that margin exists 
above the original tube design criteria. Any hypothetical accident 
as a result of any degradation in a sleeved tube would be bounded by 
the existing steam generator tube rupture accident analysis.
    Therefore, operation of the facility in accordance with proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The use of sleeves to repair degraded steam generator tubing 
will maintain the integrity of the tube bundle commensurate with the 
ASME Code and draft Regulatory Guide (RG) 1.121 margin requirements 
for original tubing. Sleeves are components which are part of the 
reactor coolant pressure boundary and meet the requirements for 
Class 1 components in Section III of the ASME Boiler and Pressure 
Vessel Code. The primary to secondary pressure boundary will be 
maintained to the same margins as the original tubes under normal 
and postulated accident conditions. The safety margins used in the 
verification of the strength of the sleeve assembly are consistent 
with the safety factors in the ASME Boiler and Pressure Vessel Code 
used in steam generator design. Further, a test program has been 
conducted by Westinghouse which demonstrated the integrity of the 
lower hard rolled joint design and its capability to withstand the 
design loads.
    Therefore, operation of the facility with 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 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, P.O. Box 19557, Irvine, California 92713.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, P. O. Box 800, Rosemead, California 91770.
    NRC Project Director: William H. Bateman.

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

    Date of application for amendments: February 26, 1999 (TS 98-08).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah (SQN) Technical Specifications (TS) by relocating 
TS 3.7.6, ``Flood Protection Plan,'' and the associated bases to the 
SQN Technical Requirements Manual (TRM). This change does not alter the 
current requirements for implementation or surveillance testing of the 
Flood Protection Plan and future revisions of this plan will require an 
evaluation in accordance with 10 CFR 50.59.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), 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 revision to the TS relocates the requirements for 
SQN flood protection without changing the current requirements. This 
administrative relocation of the requirements will not increase the 
possibility of an accident.
    The capability of the Flood Protection Plan will continue to 
provide the same function. Changes to the relocated requirements 
will be processed, in accordance with 10 CFR 50.59, to ensure the 
Flood Protection Plan will be properly maintained. Therefore, the 
proposed relocation of the flood protection requirements will not 
increase the probability or consequences of an accident previously 
evaluated.
    The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The SQN Flood Protection Plan is used to mitigate the effects of 
a flooding event at SQN. This plan would not be the initiator of any 
new or different kind of accident. The capability of the Flood 
Protection Plan will continue to provide the same function. Changes 
to the relocated requirements will be processed, in accordance with 
10 CFR 50.59, to ensure the Flood Protection Plan will be properly 
maintained. The proposed change does not alter the current functions 
of SQN's Flood Protection Plan; therefore, this proposed change will 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    B. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The requirements for SQN's flood protection are unchanged by the 
proposed relocation of the requirements to the SQN TRM. The function 
of the Flood Protection Plan and surveillance requirements to ensure 
implementation of the plan remains unchanged. Any future changes to 
these requirements will be evaluated, in accordance with 10 CFR 
50.59, to ensure acceptability and NRC review as required. 
Accordingly, the proposed change will not result in 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 
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 Project Director: Cecil O. Thomas.

[[Page 14287]]

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

    Date of application for amendments: February 26, 1999 (TS 99-02).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah (SQN) Technical Specifications (TS) to provide for 
consistency when exiting the action statements associated with the 
Emergency Diesel Generators (D/Gs). The Tennessee Valley Authority 
(TVA) inadvertently omitted revising Action Statements c, d, and e 
associated with TS 3.8.1.1 in Revision 1 to TS Change 96-08, addressing 
the D/G allowed outage time, submitted to the NRC staff on October 8, 
1998.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), TVA 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.
    This proposed revision provides for consistency and removes 
contradictions within the action statements associated with TS 
3.8.1.1. Additionally, the proposed revision will not result in any 
change in the design, maintenance or operation of the associated 
plant equipment nor will it result in deviation from the actions 
presently approved by the staff for SQN's response to the associated 
LCOs [Limiting Conditions for Operation]. The deletion of the 
defined portion of the requirements associated with the restoration 
of offsite power sources in Action Statements c and d does not 
result in any change to SQN's response to the stated condition since 
this requirement remains unchanged in Action Statement a.
    The deletion of the requirements associated with the restoration 
of 4 diesel generator (D/G) sets within 72 hours from Action 
Statements c and e provides for a consistent allowed outage time of 
7 days for the loss of a D/G set as previously approved by the staff 
in a safety evaluation issued on December 16, 1998. Therefore, the 
proposed amendment does not involve an increase in the probability 
or consequences of an 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 proposed change provides for consistency and removes 
contradictions within the action statements associated with TS 
3.8.1.1. Additionally, the proposed revision will not result in any 
change in the design, maintenance or operation of the associated 
plant equipment nor will it result in deviation from the actions 
presently approved by the staff for SQN's response to the associated 
LCOs. Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change provides for consistency and removes 
contradictions within the action statements associated with TS 
3.8.1.1. Additionally, the proposed revision will not result in any 
change in the design, maintenance or operation of the associated 
plant equipment nor will it result in deviation from the actions 
presently approved by the staff for SQN's response to the associated 
LCOs. 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 
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 Project Director: Cecil O. Thomas.

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

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

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone Nuclear 
Power Station, Unit 1, New London County, Connecticut

    Date of application of amendment: December 4, 1998, January 18, and 
January 19, 1999.
    Brief description of amendment: The proposed amendment would modify 
the staffing and training requirements to allow the use of Certified 
Fuel Handlers to meet plant staffing requirements.
    Date of publication individual notice in Federal Register: December 
29, 1998 (63 FR 71657).
    Expiration date of individual notice: January 28, 1999.
    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.

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.

[[Page 14288]]

Baltimore Gas and Electric Company, Docket No. 50-318, Calvert Cliffs 
Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

    Date of application for amendment: July 20, 1998, as supplemented 
December 4, 1998, and December 23, 1998.
    Brief description of amendment: The amendment permits a one-time 
change to the Technical Specification (TS) Bases for TS 3.8.2 for 
Calvert Cliffs Nuclear Power Plant, Unit No. 2 and provides approval of 
the licensee's analysis of unreviewed safety questions as described in 
10 CFR 50.59. The change allows Baltimore Gas and Electric Company to 
provide alternate cooling to the Unit 2 emergency diesel generators 
(EDGs) during their replacement of the Unit 2 service water (SRW) heat 
exchangers in the 1999 refueling outage since the normal SRW cooling 
would be unavailable. The licensee proposes to provide the 2A EDG with 
cooling water from the Unit 1 SRW system and to provide the 2B EDG with 
cooling water from an independent external cooling system during the 
replacement work.
    Date of issuance: March 8, 1999.
    Effective date: As of the date of its issuance to be implemented 
during the Calvert Cliffs Unit No. 2 spring 1999 refueling outage.
    Amendment No.: 205.
    Facility Operating License No. DPR-69: Amendment revised the 
Technical Specifications Bases.
    Date of initial notice in Federal Register: August 26, 1998 (63 FR 
45523) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendment: October 9, 1998.
    Brief description of amendment: The amendment revised Section 6.0 
to Technical Specifications to change the membership of the Nuclear 
Facility Safety Committee and corrected other typographical errors.
    Date of issuance: March 8, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 199.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 16, 1998 (63 
FR 69337).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: April 9, 1998 (NRC-98-0071).
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.7.1.2, ``Emergency Equipment Cooling Water 
System,'' Action a, and TS 3.8.1.1, ``A.C. Sources--Operating,'' Action 
c, to be consistent with the actions required for inoperable oxygen 
monitoring instrumentation in TS 3.3.7.5, ``Accident Monitoring 
Instrumentation.'' The existing ``**'' footnote to TS 3.7.1.2, Action 
a, is modified and a ``*'' footnote is added to TS 3.8.1.1, Action c.
    Date of issuance: March 3, 1999.
    Effective date: March 3, 1999, with full implementation within 30 
days.
    Amendment No.: 132.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 23, 1998 (63 
FR 50937).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 3, 1999.
    No significant hazards consideration comments received: No.
     Local Public Document Room location: Monroe County Library System, 
Ellis Reference and Information Center, 3700 South Custer Road, Monroe, 
Michigan 48161.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: September 23, 1998.
    Brief description of amendment: The amendment changes Division III 
battery specific gravity acceptance criteria outlined in River Bend 
Station (RBS) Technical Specifications (TS). The change is required as 
a result of Division III battery system modifications scheduled to be 
implemented during refueling outage RF-8, beginning April 3, 1999. 
During this time, the current Division III battery will be replaced 
with a new battery having a greater capacity rating. The new battery 
has a nominal specific gravity of 1.215 at 77 deg.F in contrast to the 
existing Division III battery supplied with a nominal specific gravity 
of 1.210 at 77 deg.F. Since TS Section 3.8.6, Table 3.8.6-1 values for 
specific gravity are based upon the manufacturer's nominal specific 
gravity, these values were updated to reflect the changes.
    Date of issuance: March 3, 1999.
    Effective date: The license amendment is effective upon the date of 
issuance and shall be implemented within 90 days.
    Amendment No.: 103.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 64111).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803.

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

    Date of application for amendment: August 31, 1998.
    Brief description of amendment: This amendment revised Technical 
Specification Surveillance Requirement 3.6.1.3.4 to permit removal of 
the inclined fuel transfer system primary containment blind flange 
while primary containment integrity is required.
    Date of issuance: February 24, 1999.
    Effective date: February 24, 1999.
    Amendment No.: 100.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 21, 1998 (63 FR 
56260).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 24, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081.

[[Page 14289]]

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

    Date of application for amendment: July 13, 1998, and as 
supplemented by submittal dated November 23, 1998.
    Brief description of amendment: This amendment revised Technical 
Specification 3.4.4,'' Safety/Relief Valves (SRVs),'' by increasing the 
present plus or minus 1% tolerance on the safety mode lift setpoint for 
the safety relief valves to plus or minus 3%.
    Date of issuance: March 3, 1999.
    Effective date: March 3, 1999.
    Amendment No.: 101.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 12, 1998 (63 FR 
43214).
    The supplemental information contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081.

Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit 
1, DeWitt County, Illinois

    Date of application for amendment: October 5, 1998.
    Brief description of amendment: The amendment allows deferral of 
the next scheduled local leak rate test for valve 1MC-042 until the 
seventh refueling outage.
    Date of issuance: March 8, 1999.
    Effective date: March 8, 1999, and shall be implemented within 45 
days.
    Amendment No.: 121.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 23, 1998 (63 FR 
56949).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, IL 61727.

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone Nuclear 
Power Station, Unit 1, New London County, Connecticut

    Date of application for amendment: December 4, 1998, and January 18 
and 19, 1999.
    Brief description of amendment: The proposed amendment would modify 
the staffing and training requirements to allow the use of Certified 
Fuel Handlers to meet plant staffing requirements.
    Date of issuance: March 5, 1999.
    Effective date: As of the date of issuance to be implemented within 
45 days from the date of issuance.
    Amendment No.: 104.
    Facility Operating License No. DPR-21: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 29, 1998 (63 
FR 71657).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 5, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

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

    Date of applications for amendment: August 12, 1998, as 
supplemented by letter dated October 30, 1998; and application dated 
September 28, 1998, as supplemented by letters dated January 7 and 20, 
1999.
    Brief description of amendment: The amendment allows implementation 
of a revised main steamline break analysis and revised control room 
habitability analyses.
    Date of issuance: March 10, 1999.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 228.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications and authorized changes to the Final Safety 
Analysis Report.
    Date of initial notice in Federal Register: October 7, 1998 (63 FR 
53951) and December 2, 1998 (63 FR 66597).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 10, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

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

    Date of application for amendment: December 10, 1998, as 
supplemented February 19, 1999.
    Brief description of amendment: The amendment allows the licensee 
to implement changes to the Final Safety Analysis Report (FSAR) 
regarding a revised method for ensuring boron precipitation can be 
prevented (post-loss-of-coolant accident).
    Date of issuance: March 10, 1999.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 229.
    Facility Operating License No. DPR-65: Amendment authorizes changes 
to the Final Safety Analysis Report.
    Date of initial notice in Federal Register: January 13, 1999 (64 FR 
2249).
    The February 19, 1999, supplemental letter provided additional 
information that did not change the staff's proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 10, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

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

    Date of application for amendment: June 10, 1998, as supplemented 
October 30, 1998.
    Brief description of amendment: The amendment revises the Millstone 
Unit 3 licensing basis associated with post-accident mitigation 
activities, vital area access travel routes, and the associated action 
completion times. Northeast Nuclear Energy Company determined that the 
Final Safety Analysis Report (FSAR) description of post-accident vital 
area routing was out of date

[[Page 14290]]

because the radiological control area boundary fence created an access 
problem on the designated routes to the hydrogen recombiner and fuel 
building. The revised licensing basis will be incorporated into the 
FSAR and will revise the routes to accommodate the fence location and 
allow for the time to unlock gates.
    Date of issuance: March 1, 1999.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 166.
    Facility Operating License No. NPF-49: Amendment authorized 
revision to the FSAR.
    Date of initial notice in Federal Register: July 15, 1998 (63 FR 
38202).
    The October 30, 1998, letter provided clarifying information that 
did not change the scope of the June 10, 1998, application, and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 1, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

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

    Date of application for amendment: December 4, 1998.
    Brief description of amendment: The amendment eliminates the need 
to cycle the plant and its components through a shutdown-startup cycle 
by allowing the next snubber surveillance interval to be deferred until 
the end of refueling outage 6 or September 10, 1999, whichever date is 
earlier.
    Date of issuance: March 3, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 167.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 30, 1998 (63 
FR 71971).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
ThreeRivers 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 No. 50-388, Susquehanna Steam Electric Station, Unit 
2, Luzerne County, Pennsylvania

    Date of application for amendment: August 5, 1998, as supplemented 
by letter dated November 23, 1998.
    Brief description of amendment: This amendment would change the 
allowable values for both the core spray system and the low-pressure-
coolant injection system reactor steam dome pressure-low functions.
    Date of issuance: March 4, 1999.
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment No.: 155.
    Facility Operating License No. NPF-22: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 1, 1999 (64 FR 
4904).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 4, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

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

    Date of application for amendment: August 25, 1998, as supplemented 
January 27, 1999.
    Brief description of amendment: This amendment revised Technical 
Specification (TS) 2.1.2, ``THERMAL POWER, High Pressure and High 
Flow,'' and the Bases for TS 2.1, ``Safety Limits.'' These changes were 
made to implement appropriately conservative Safety Limit Minimum 
Critical Power Ratio values for the Hope Creek Generating Station Cycle 
9 core and fuel designs. An administrative revision has also been made 
to TS 6.9.1.9 to reflect these changes for Cycle 9.
    Date of issuance: March 9, 1999.
    Effective date: As of the date of issuance, to be implemented 
within 60 days after the completion of Cycle 8.
    Amendment No.: 117.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 23, 1998 (63 
FR 50938).
    The supplemental letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 9, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.

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

    Date of application for amendments: September 29, 1998.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3/4.9.4, ``Refueling Operations--Containment 
Building Penetrations,'' to allow the use of an equivalent closure 
device to satisfy the closure requirements of the containment equipment 
hatch during core alterations or movement of irradiated fuel in 
containment. The amendment also revises TS 3/4.9.4 to allow the use of 
an equivalent closure method to satisfy the closure requirements of 
containment penetrations (in addition to an isolation valve, blind 
flange or manual valve) during core alterations or movement of 
irradiated fuel in containment.
    Date of issuance: February 26, 1999.
    Effective date: Effective as of its date of issuance, to be 
implemented within 60 days.
    Amendment Nos.: 217 and 199.
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 21, 1998 (63 FR 
56258).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

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

    Date of application for amendment: November 24, 1998.
    Brief description of amendment: This amendment revises the Ginna 
Station Improved Technical Specifications

[[Page 14291]]

description of the fuel cladding material (TS 4.2.1) and updates the 
list of references provided in Specification 5.6.5 for the Core 
Operating Limits Report.
    Date of issuance: March 3, 1999.
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment No.: 73.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 30, 1998 (63 
FR 71972).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: July 7, 1998, as supplemented by letters 
dated October 15 and October 26, 1998, and February 16, 1999. The 
supplements provided clarifying information and corrected 
administrative errors within the scope of the amendment request and did 
not change the initial no significant hazards consideration 
determination.
    Brief description of amendments: The amendments revised the spent 
fuel pool criticality analysis and rack utilization schemes by allowing 
credit for spent fuel pool soluble boron.
    Date of issuance: March 3, 1999.
    Effective date: This license amendment is effective as of its date 
of issuance and shall be implemented within 90 days of issuance.
    Amendment Nos.: Unit 1--Amendment No. 104; Unit 2--Amendment No. 
91.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 26, 1998 (63 FR 
45530).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear 
Plant, Unit 2, Limestone County, Alabama

    Date of application for amendment: September 8, 1998 (TS-354), as 
supplemented by letter dated February 22, 1999.
    Brief description of amendment: Revises the Appendix A Technical 
Specifications (TS) to include provisions for enabling the Oscillation 
Power Range Monitor Upscale trip function in the Average Power Range 
Monitor.
    Date of issuance: As of date of issuance to be implemented at the 
end of the Unit 2 Cycle 10 outage scheduled to begin on April 11, 1999.
    Effective date: March 5, 1999.
    Amendment No.: 258.
    Facility Operating License No. DPR-52: Amendment revises the TS.
    Date of initial notice in Federal Register: October 7, 1998 (63 FR 
53958). The supplemented letter dated February 22, 1999, did not change 
the original no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 5, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611.

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of application for amendments: November 10, 1998.
    Brief description of amendments: The amendments revise the 
Technical Specifications Sections 3.4.4 and 3.4.4.a for Unit 1, and 
3.4.4 and 3.4.4.a for Unit 2, providing a clarification on the 
operability requirements for pressurizer heaters and the emergency 
power source for the pressurizer heaters.
    Date of issuance: March 1, 1999.
    Effective date: March 1, 1999.
    Amendment Nos.: 217 and 198.
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: December 2, 1998 (63 FR 
66605).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 1, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of application for amendments: October 25, 1995, as 
supplemented February 5, 1999. The February 5, 1999, supplemental 
letter contained clarifying information only, and did not change the 
initial no significant hazards consideration determination or expand 
the scope of the original Federal Register Notice.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) Sections 3.4.3.2, 4.4.3.2.1.b, 
4.4.3.2.1.c, 4.4.3.2.2, 4.4.9.3.d, 4.4.9.3.e, 3/4.4.2, 3/4.4.3, 3/4.4.4 
and 6.8.4.g for Unit 1, and 3.4.3.2, 4.4.3.2.1.c, 4.4.3.2.2, 4.4.9.3.d, 
4.4.9.3.e, 3/4.4.2, 3/4.4.3, 3/4.4.4 and 6.8.4.g for Unit 2, providing 
an allowed outage time of 14 days for the prezzurizer power operated 
relief valve (PORV) nitrogen accumulators, as well as provide separate 
action statements for the PORV depending on the reason for the PORV 
inoperability.
    Date of issuance: March 2, 1999.
    Effective date: March 2, 1999.
    Amendment Nos.: 218 and 199.
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28620).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 2, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

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

    Date of application for amendments: September 24, 1998.
    Brief description of amendments: These amendments revise the 
Technical Specifications to allow the reactor trip bypass breakers to 
be tested immediately after being placed in service, but prior to 
commencing Reactor Protection System testing or maintenance.
    Date of issuance: March 12, 1999.
    Effective date: March 12, 1999.
    Amendment Nos.: 219 and 219.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
change the Technical Specifications.

[[Page 14292]]

    Date of initial notice in Federal Register: February 10, 1999 (64 
FR 6715).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 12, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.

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

    Date of application for amendments: September 28, 1998 (TSCR 208).
    Brief description of amendments: These amendments clarify the 
notation definition of refueling interval ``R'' in TS Table 15.4.1-1 
and add a new annual (12-month) interval ``A''.
    Date of issuance: March 1, 1999.
    Effective date: March 1, 1999, with full implementation within 45 
days.
    Amendment Nos.: 186 and 191.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 27, 1999 (64 FR 
4162).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 1, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.

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

    Date of application for amendments: October 5, 1998 (TSCR 200).
    Brief description of amendments: These amendments modify TS Section 
15.4.1, ``Operational Safety Review,'' by removing the requirement to 
check certain environmental monitors on a monthly basis.
    Date of issuance: March 2, 1999.
    Effective date: March 2, 1999, with full implementation within 45 
days.
    Amendment Nos.: 187 and 192.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 27, 1999 (64 FR 
4163).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 2, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.

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

    Date of application for amendments: October 7, 1998 (TSCR 207).
    Brief description of amendments: These amendments incorporate 
changes to the Technical Specifications to ensure the 4 kV bus 
undervoltage input to the reactor trip protective function is 
controlled in accordance with the design and licensing basis for the 
facility. An additional administrative change removes the footnote 
related to the definition of Rated Power in TS 15.1.j.
    Date of issuance: March 2, 1999.
    Effective date: March 2, 1999, with full implementation within 45 
days.
    Amendment Nos.: 188 and 193.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 30, 1998 (63 
FR 71978).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 2, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.

Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power 
Station, Franklin County, Massachusetts

    Date of application for amendment: October 15, 1998.
    Brief description of amendment: Revises the Possession Only License 
by changing the submittal interval for the Radioactive Effluent Reports 
from semiannual to annual.
    Date of issuance: March 5, 1999.
    Effective date: March 5, 1999.
    Amendment No.: 151.
    Possession Only License No. DPR-3: Amendment revised the Technical 
Specifications.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 64128). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Greenfield Community College, 
1 College Drive, Greenfield, Massachusetts 01301.

    Dated at Rockville, Maryland, this 17th day of March 1999.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 99-7032 Filed 3-23-99; 8:45 am]
BILLING CODE 7590-01-P