[Federal Register Volume 64, Number 173 (Wednesday, September 8, 1999)]
[Notices]
[Pages 48858-48875]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-23300]


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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 August 14, 1999, through August 27, 1999. 
The last biweekly notice was published on August 25, 1999 (64 FR 
46424).

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.

[[Page 48859]]

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 October 8, 1999, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendment request: August 3, 1999.

[[Page 48860]]

    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 2.1.B to increase the minimum 
critical power ratio for higher cycle exposures for Unit 2. The 
proposed amendments would also revise TS 6.9.A.6.b for Units 2 and 3 to 
add an NRC-approved topical report to the list of analytical 
methodologies that are used to determine operating limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. Limits have been established consistent with NRC-
approved methods to ensure that fuel performance during normal, 
transient, and accident conditions is acceptable. These changes do 
not affect the operability of plant systems, nor do they compromise 
any fuel performance limits.
    Changing the Minimum Critical Power Ratio (MCPR) Safety Limit 
(SL) at Dresden Nuclear Power Station Unit 2 will not increase the 
probability or the consequences of an accident previously evaluated. 
This change implements the MCPR SL resulting from the Siemens Power 
Corporation (SPC) ANFB critical power correlation methodology using 
the approved ATRIUM-9B additive constant uncertainty. For each 
cycle, specific MCPR SL calculations will be performed, consistent 
with SPC's approved methodology, to confirm the appropriateness of 
the MCPR SL. Additionally, operational MCPR limits will be applied 
that will ensure the MCPR SL is not violated during all modes of 
operation and anticipated operational occurrences. The MCPR SL 
ensures that less than 0.1% of the rods in the core are expected to 
experience boiling transition. Therefore, the probability or 
consequences of an accident will not increase.
    Adding EMF-85-74, Revision 0, Supplements 1 and 2 (P)(A) to 
Section 6 for Dresden Nuclear Power Station Units 2 and 3, does not 
increase the probability or consequences of an accident previously 
evaluated. The NRC-approved burnup extension for RODEX2A 
applications has been demonstrated to meet all applicable design 
criteria. Therefore, adding this methodology to Technical 
Specification Section 6 does not increase to the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated:
    Creation of the possibility of a new or different kind of 
accident would require the creation of one or more new precursors of 
that accident. New accident precursors may be created by 
modifications to the plant configuration, including changes in 
allowable modes of operation. This Technical Specification submittal 
does not involve any modifications to the plant configuration or 
allowable modes of operation. No new precursors of an accident are 
created and no new or different kinds of accidents are created. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Changing the MCPR SL does not create the possibility of a new 
accident from any accident previously evaluated. This change does 
not alter or add any new equipment or change modes of operation. The 
MCPR SL is established to ensure that 99.9% of the rods avoid 
boiling transition.
    The MCPR SL is changing for Dresden Nuclear Power Station Unit 2 
to support Cycle 17 operation. This change does not introduce any 
physical changes to the plant, the processes used to operate the 
plant, or allowable modes of operation. Therefore, no new accidents 
are created that are different from any accident previously 
evaluated.
    The addition of RODEX2A (EMF-85-74, Revision 0, Supplements 1 
and 2 (P)(A)) to Section 6 does not create the possibility of a new 
accident from an accident previously evaluated. This change does not 
alter or add any new equipment or change modes of operation. This 
change does not introduce any physical changes to the plant, the 
processes used to operate the plant, or allowable modes of 
operation. Therefore, no new accidents are created that are 
different from any accident previously evaluated.
    3. Involve a significant reduction in the margin of safety for 
the following reasons:
    Changing the MCPR SL for Dresden Nuclear Power Station Unit 2 
will not involve any reduction in margin of safety. The MCPR SL 
provides a margin of safety by ensuring that less than 0.1% of the 
rods are calculated to be in boiling transition. The proposed 
Technical Specification amendment request reflects the MCPR SL 
results from evaluations by SPC using NRC-approved methodology.
    Because the methodology used to determine the MCPR SL is 
conservative and has received NRC approval, a decrease in the margin 
to safety will not occur due to changing the MCPR SL. The revised 
MCPR SL will ensure the appropriate level of fuel protection. 
Additionally, operational limits will be established based on the 
proposed MCPR SL to ensure that the MCPR SL is not violated during 
all applicable modes of operation including anticipated operation 
occurrences. This will ensure that the fuel design safety criterion 
of more than 99.9% of the fuel rods avoiding transition boiling 
during normal operation as well as during an anticipated operational 
occurrence is met.
    The addition of EMF-85-74, Revision 0, Supplements 1 and 2 
(P)(A) to Section 6 does not decrease the margin of safety. The 
burnup limit extension for RODEX2A applications has been reviewed 
and approved by the NRC. The data supporting the burnup extension 
demonstrates that all applicable design criteria are met. Therefore, 
since the burnup extension is acceptable and within the design 
criteria, using the approved burnup extension will not affect the 
margin of safety.

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

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

    Date of amendment request: August 13, 1999, as supplemented on 
August 27, 1999.
    Description of amendment request: The proposed amendments would 
revise Technical Specification Section 1.0, ``Definitions,'' Item 1.7, 
``Core Alteration,'' to specify that movement of instrumentation and 
control rod movements are not considered core alterations if there are 
no fuel assemblies in the associated cell. The licensee also proposed 
corresponding changes to TS Sections 3/4.1, 3/4.3, and 3/4.9 to reflect 
the change in definition.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes incorporate a definition contained in 
NUREG-1433, Revision 1, ``Standard Technical Specifications, General 
Electric Plants, BWR/4.'' There are no modifications to plant 
equipment or systems and there is no direct effect on plant 
operation. The proposed changes do not affect any accident 
initiators or precursors and do not change or alter the design 
assumptions for systems or components used to mitigate the 
consequences of an accident. The proposed changes do not affect the 
design or operation of any system, structure, or component in the 
plant. The proposed changes do not impact

[[Page 48861]]

the requirements for refueling evolutions associated with shutdown 
margin, core monitoring, and reactor protection system operability. 
There are no changes to parameters governing plant operation, and no 
new or different types of equipment will be installed. These changes 
do not impact any accident previously evaluated in the Updated Final 
Safety Analysis Report (UFSAR). Therefore, no increases in the 
probability of an accident or consequences will result due to this 
change.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes do not affect the design or operation of 
any plant system, structure, or component. There are no changes to 
parameters governing plant operation, and no new or different type 
of equipment will be installed. There is no change in any method by 
which a safety related system performs its function. No new 
equipment is being introduced, and installed equipment is not being 
operated in a new or different manner. There are no setpoints 
affected by this proposed action. This proposed action will not 
alter the manner in which equipment operation is initiated, nor will 
the function demands on credited equipment be changed. As such, no 
new failure modes are being introduced. There are no changes to 
assumptions in accident analysis. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The proposed changes are consistent with NUREG-1433, Revision 1, 
``Standard Technical Specifications, General Electric Plants, BWR/
4.'' The proposed changes do not adversely affect existing plant 
safety margins or the reliability of the equipment assumed to 
operate in the safety analysis. The initial conditions and 
methodologies used in the accident analyses remain unchanged. 
Therefore, accident analyses results are not impacted. There are no 
resulting effects on plant safety parameters or setpoints. The 
proposal does not involve a significant relaxation of the criteria 
used to establish safety limits, a significant relaxation of the 
bases for the limiting safety system settings, or a significant 
relaxation of the bases for the limiting conditions for operations. 
Therefore, these proposed changes do not cause a reduction in the 
margin of safety.

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

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

    Date of amendment request: May 5, 1999.
    Description of amendment request: The proposed amendment would 
permit a one-time extension of the allowed outage time (AOT) for the 
reactor protection and engineered safety feature actuation 
instrumentation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The reactor protection and engineered safety features functions 
are not initiators of any design basis accident or event and 
therefore do not increase the probability of any accident previously 
evaluated. The proposed changes to the AOTs, bypass times, and 
allowing on-line testing and maintenance have an insignificant 
impact on plant safety based on the calculated CDF [core damage 
frequency] increase being less than LOE-06. Therefore, the proposed 
changes do not result in 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 changes do not result in a change in the manner in 
which the RPS [reactor protection system] and ESFAS [engineered 
safety features actuation system] provide plant protection. No 
change is being made which alters the functioning of the RPS and 
ESFAS. Rather, the likelihood or probability of the RPS or ESF 
functioning properly is affected as described above. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident nor involve a reduction in the margin of safety as 
defined in the Safety Analysis Report.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system setpoints or limiting conditions for 
operations are determined. The impact of increased AOTs, testing 
times, and allowing on-line testing and maintenance are expected to 
result in an overall improvement in safety because:
    The longer AOTs for the master relays, logic cabinets, and 
analog channels will promote improved maintenance practices that 
will provide improved component performance, improved availability 
of the protection system, and a reduced number of spurious reactor 
trips and spurious actuation of safety equipment.
    The longer AOTs and bypass times for the analog channels will 
provide additional time before being required to place the channel 
in trip. With the channel in trip, the logic required to cause a 
reactor trip or a safety system actuation is reduced to 1 of 2 (for 
2 of 3 logic) and to 1 of 3 (for 2 of 4 logic). With the reduced 
logic requirement, the potential for a spurious actuation is 
increased. Leaving the channel in the bypass state for additional 
time does reduce the availability of signals to initiate component 
actuation for event mitigation when required, but as shown in this 
analysis, the impact on plant safety is small due to the 
availability of other signals or operator action to trip the reactor 
or cause component actuation.
    The longer allowed outage times will provide plant operators 
additional flexibility in operating the plant. There will be 
additional time available before an action needs to be taken to shut 
down the plant or place a channel in the tripped state. This 
additional flexibility will facilitate prioritizing component 
repairs.

    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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610. Biweekly Notice 
Coordinator Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving 
Place, New York, New York 10003.
    NRC Section Chief: S. Singh Bajwa.

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

    Date of amendment request: August 4, 1999.
    Description of amendment request: The amendments would revise the 
joint Technical Specifications as follows:
    (1) A current action in Section 3.2.2 requires that when one 
Nuclear Service Water System (NSWS) suction transfer low pit level 
channel is inoperable, the channel be placed in its trip position. The 
licensee proposed an additional alternative such that the NSWS suction 
can simply be aligned from Lake Wylie to the Standby Nuclear Service 
Water Pond (SNSWP). Suction from Lake Wylie is the normal 
configuration, while suction from the SNSWP is the safety 
configuration. This proposed alternative

[[Page 48862]]

action provides operational flexibility; there is no associated design 
change to the units.
    (2) The licensee proposed to delete from Table 3.3.2-1, 
``Engineered Safety Feature Actuation System Instrumentation,'' the 
entry regarding Auxiliary Feedwater Loss of Offsite Power (Function 6d) 
on the basis that a comparable and adequate requirement will exist in 
Section 3.3.5. To such end, a new Surveillance Requirement (SR) 3.3.5.3 
will be added, incorporating the Function 6d requirement from Table 
3.3.2-1. These proposed changes remove inconsistencies that currently 
exist in the Technical Specifications for Function 6d. There is no 
associated design change to the units.
    (3) In the process of converting the Technical Specification to the 
improved format (Amendment Nos.173 and 165), errors were inadvertently 
introduced regarding the conditions under which the Reactor Coolant 
System Subcooling Margin Monitor must be operable. The licensee 
proposed to correct these errors by revising the entry regarding the 
Subcooling Margin Monitor in Table 3.3.3-1, ``Post Accident Monitoring 
Instrumentation''. There is no associated design change to the units.
    (4) Section 3.4.17 is concerned with reactor coolant system loops 
test exceptions. Currently Surveillance Requirement 3.4.17.2 
incorrectly specifies that a COT [channel operational test] be 
performed ``for each power range neutron flux-flow and intermediate 
range neutron flux channel and P-7 [Low Power Reactor Trips Block 
Function]''. The licensee proposed to correct this statement by 
deleting ``P-7'' and adding ``P-10 [Power Range Neutron Flux] and P-13 
[Turbine Impulse Pressure]''. This correction does not involve any 
design change to the units.
    (5) The licensee proposed to delete from Section 5.3.1 the specific 
qualification requirements for Reactor Operators (ROs) and Senior 
Reactor Operators (SROs). Such requirements are specified by 10 CFR 
50.55, ``Operators'' Licenses'', and the licensee is required to follow 
this regulation. There will be no change in the qualification of ROs 
and SROs, and no design change to the units.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

First Standard

    Implementation of this amendment would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. Approval of this amendment will have no effect 
on accident probabilities or consequences. For proposed changes #1-
4, the systems and equipment referenced in the revised TS are not 
accident initiating systems; therefore, there will be no impact on 
any accident probabilities by the approval of this amendment. The 
design of the systems is not being modified by these proposed 
changes. Therefore, there will be no impact on any accident 
consequences. For proposed change #5, the change is purely 
administrative; it will therefore have no effect on any accident 
probabilities or consequences.

Second Standard

    Implementation of this amendment would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. No new accident causal mechanisms are created 
as a result of NRC approval of this amendment request. No changes 
are being made to the plant which will introduce any new accident 
causal mechanisms. This amendment request does not impact any plant 
systems that are accident initiators; neither does it adversely 
impact any accident mitigating systems.

Third Standard

    Implementation of this amendment would not involve a significant 
reduction in a margin of safety. Margin of safety is related to the 
confidence in the ability of the fission product barriers to perform 
their design functions during and following an accident situation. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment system. The performance of these fission 
product barriers will not be impacted by implementation of this 
proposed amendment. The systems and equipment referenced in the 
revised TS for proposed changes #1-4 are already capable of 
performing as designed. No safety margins will be impacted. Since 
proposed change #5 is purely administrative, it will have no effect 
on any safety margins.

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

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

    Date of amendment request: July 26, 1999.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) Section 3/4.3.2.1, ``Safety 
Features Actuation System Instrumentation,'' to remove the ``Trip 
Setpoint'' values and revise the ``Allowable Values'' entries for 
Sequence Logic Channels a, ``Essential Bus Feeder Breaker Trip (90%),'' 
and b, ``Diesel Generator Start, Load Shed on Essential Bus (59%).''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the 
proposed changes and determined that a significant hazards 
consideration does not exist because operation of the Davis-Besse 
Nuclear Power Station, Unit No. 1, in accordance with these changes 
would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because the proposed changes do not 
change any accident initiator, initiating condition, or assumption.
    The proposed changes would revise Technical Specification (TS) 
Table 3.3-4, Safety Features Actuation System Instrumentation Trip 
Setpoints, to remove the'Trip Setpoint'' values for Functional Unit 
Sequence Logic Channel ``a'', ``Essential Bus Feeder Breaker Trip 
(90%)'', and Functional Unit Sequence Logic Channel ``b'', ``Diesel 
Generator Start, Load Shed on Essential Bus (59%)'', and also modify 
the ``Allowable Values'' entry for Functional Unit Sequence Logic 
Channel ``a'', consistent with updated calculations and current 
setpoint methodology. The proposed changes would also clarify an 
inconsistency between Table 3.3-4 and Table 4.3-2, Safety Features 
Actuation System Instrumentation Surveillance Requirements. The 
proposed changes to Limiting Condition for Operation (LCO) 3.3.2.1 
and Bases 3/4.3.1 and 3/4.3.2 are associated with these changes.
    The accident previously evaluated in Section 15.2.9, ``Loss of 
All AC Power to the Station Auxiliaries (Station Blackout),'' of the 
DBNPS Updated Safety Analysis Report (USAR) is not affected by the 
proposed changes because its bounding conditions are not affected. 
The existing TS action statements will continue to maintain the USAR 
requirement to start and load one Emergency Diesel Generator (EDG) 
to meet minimum ESF requirements, should all AC power be lost. 
Furthermore, the proposed changes are based on the existing 
performance characteristics of plant equipment; therefore, the 
proposed changes

[[Page 48863]]

will not involve a significant change to the plant design or 
operation.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
invalidate assumptions used in evaluating the radiological 
consequences of an accident, do not alter the source term or 
containment isolation, and do not provide a new radiation release 
path or alter radiological consequences.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because the proposed 
changes do not introduce a new or different accident initiator or 
introduce a new or different equipment failure mode or mechanism.
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes do not significantly reduce the ability 
of the plant to respond to a loss of AC power to the essential 4160 
Volt buses in a timely manner. The revised Allowable Value for the 
Sequence Logic Channel ``Essential Bus Feeder Breaker Trip (90%)'' 
takes into account the need not only to be able to actuate 
Engineered Safety Features equipment coincident with a degraded grid 
condition, but to provide voltage at the required value to properly 
operate the equipment.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: July 27, 1999.
    Description of amendment request: The proposed amendment would 
remove Technical Specification (TS) Section 6.4, ``Training,'' relocate 
TS Sections 6.5.2.8, ``Audits,'' and 6.10 ``Record Retention,'' to the 
Updated Safety Analysis Report, and make related changes to TS Sections 
6.14, ``Process Control Program,'' and 6.15, ``Offsite Dose Calculation 
Manual.'' In addition, an editorial correction is proposed to TS 6.8, 
``Procedures and Programs.''
    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 Davis-Besse Nuclear Power Station has reviewed the proposed 
changes and determined that a significant hazards consideration does 
not exist because operation of the Davis-Besse Nuclear Power 
Station, Unit Number 1, in accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators, 
conditions or assumptions are affected by the proposed changes to 
Section 6.0, Administrative Controls, of the Technical 
Specifications (TS).
    The proposed changes to remove Section 6.4, Training, from the 
TS and relocate the detailed listings of TS Section 6.5.2.8, Audits, 
and TS Section 6.10, Record Retention, to the DBNPS [Davis-Besse 
Nuclear Power Station] Quality Assurance Program in Chapter 17 of 
the Updated Safety Analysis Report are consistent with NUREG-1430, 
``Standard Technical Specifications--Babcock and Wilcox Plants,'' 
Revision 1 or NRC Administrative Letter 95-06 ``Relocation of 
Technical Specification Administrative Controls Related to Quality 
Assurance,'' dated December 12, 1995. The proposed changes to TS 
Section 6.14, Process Control Program (PCP); TS Section 6.15, 
Offsite Dose Calculation Manual (ODCM); and TS Section 6.8, 
Procedures and Programs, are either associated administratively with 
the above proposed changes or are editorial corrections. These TS 
being removed or relocated will remain subject to the controls of 
regulations (e.g., 10 CFR 50.59, 10 CFR 55.59, or 10 CFR 50.54(a)).
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because no accident conditions or 
assumptions are affected by the proposed changes. As described 
above, these changes are consistent with the improved ``Standard 
Technical Specifications--Babcock and Wilcox Plants'' (NUREG-1430) 
or Administrative Letter 95-06 and are administrative changes. The 
proposed changes do not alter the source term, containment 
isolation, or allowable releases. The proposed changes, therefore, 
will not increase the radiological consequences of a previously 
evaluated accident.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new 
accident initiators or assumptions are introduced by the proposed 
changes, which involve only administrative controls. The proposed 
changes do not alter any accident scenarios.
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes are administrative and do not reduce or 
adversely affect the capabilities of any plant structures, systems 
or components to perform their nuclear safety function.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: August 5, 1999.
    Description of amendment request: The requested changes correct 
editorial errors in Technical Specification (TS) Sections 3.8.3.2, 
4.6.2.1, 4.6.2.2, 4.8.1.1, and 4.9.12. Also, the requested changes 
correct minor editorial and reference errors in Technical Specification 
Bases Sections B 3/4.3.2, B 3/4.4.11, B 3/4.6.1.2, and B 3/4.8.4.
    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:

    NNECO [Northeast Nuclear Energy Company] has reviewed the 
proposed revision in accordance with 10CFR50.92 and has concluded 
that the revision does not involve any Significant Hazards 
Considerations (SHC). The basis for this conclusion is that the 
three criteria of 10CFR50.92(c) are not satisfied. The proposed 
Technical Specification revision does not involve an SHC because the 
revision would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed TS changes are editorial in nature and do not alter 
or effect the design, operation, maintenance[,] or surveillance 
associated with MP-3 [Millstone Nuclear Power Station, Unit No. 3] 
[s]tructures, [s]ystems, and [c]omponents (SSC) during normal or 
accident operations. Since the SS[Cs] are not altered[,] 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 accident previously evaluated.
    The proposed TS changes are editorial in nature and do not alter 
or effect the design,

[[Page 48864]]

operation, maintenance[,] or surveillance associated with MP-3 
[s]tructures, [s]ystems, and [c]omponents (SSC) during normal or 
accident operations. Since the Units SS[Cs] have not been modified 
physically, or operationally[,] due to procedure changes prompted by 
this TSCR [Technical Specification Change Request], the proposed 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    These proposed TS changes are editorial and do not impact any 
MP-3 design or operational requirements. MP-3 system performance and 
operating limits are not affected; therefore[,] the proposed change 
does not involve a significant reduction in the margin of safety.
    In conclusion, based on the information provided, it is 
determined [by NNECO] that the proposed revision does not involve 
a[n] SHC.

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

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

    Date of amendment request: June 22, 1999.
    Description of amendment request: The Limerick Generating Station 
(LGS), Units 1 and 2, Technical Specifications (TS) contained in 
Appendix A to the Operating Licenses would be amended to eliminate a 
surveillance requirement for the Reactor Recirculation System. This 
proposed TS change request involves revising the TS to delete 
Surveillance Requirement 4.4.1.1.2, and associated TS Administrative 
Controls Section 6.9.1.9.h, which requires that each Reactor 
Recirculation System pump motor generator (MG) set scoop tube 
mechanical and electrical stop be demonstrated OPERABLE with the 
overspeed setpoints less than or equal to the setpoints as noted in the 
Core Operating Limits Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. The proposed Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated. The proposed TS changes do not make any 
physical changes to the fuel, or the way the fuel responds to a 
transient or accident. The radiological barriers are not compromised. 
The fuel will continue to be operated to analyzed operating limits. No 
new failure mode is introduced.
    Prior to the removal of the Recirculation System Master Flow 
Controller at LGS, the bounding postulated event involving an increase 
in reactor coolant system flow rate was the dual pump slow flow runout 
event not terminated by SCRAM. The requirements surrounding the MG set 
stops were established to mitigate consequences during a dual pump slow 
flow runout by providing a limit on the maximum core flow. The MG set 
stop requirements were not established to prevent an accident. The 
potential common mode failure required for a dual pump slow flow runout 
event was eliminated with the removal of the Master Flow Controller. 
The elimination of the Master Flow Controller does not increase the 
probability of other core flow increase events, or of any other events 
previously analyzed.
    Revised generic flow biased ARTS [APRM (average power range 
monitor)/RBM (rod block monitor) Technical Specifications Improvement] 
thermal limits that do not take credit for MG set stops have been 
developed for LGS, Units 1 and 2. Adherence to approved flow biased 
ARTS thermal limits identified in the LGS, Units 1 and 2, Core 
Operating Limits Reports (COLRs) ensure that fuel design limits are not 
exceeded. Maintaining fuel design limits results in no change in the 
consequences of accidents previously evaluated.
    The single pump slow flow runout does not terminate by Main Steam 
Isolation Valve (MSIV) closure or generator load reject. As a result, 
the single pump runout event does not result in any significant 
pressurization and does not represent a challenge to the reactor 
coolant pressure boundary. MSIV closure with associated SCRAM on high 
neutron flux, as confirmed in the cycle specific Supplemental Reload 
Licensing Report (SRLR), remains the bounding reactor pressure vessel 
overpressurization event for LGS, Units 1 and 2. In addition, there are 
no other associated impacts to the plant resulting from a single pump 
runout. Therefore, the integrity of radiological barriers will not be 
compromised.
    Although there is no longer a safety need to demonstrate 
operability of the MG set stops, there still is an operational need to 
have the MG set stops for the Reactor Recirculation System (RS). Damage 
to the jet pump sensing lines could occur if the resonance frequency of 
the sensing lines is reached. Jet pump sensing line tests established a 
conservative pump speed limit (1650 rpm for Unit 1, no limit for Unit 
2) to preclude sensing line resonance. The MG set stop setpoint bounded 
the operationally required setpoint. The operationally required MG set 
stop setpoint to preclude jet pump sensing line resonance will continue 
to be controlled administratively via approved plant procedures. The 
proposed TS changes do not adversely impact the RS, or introduce new or 
unanalyzed operating conditions for the RS. The MG sets will not exceed 
their previously analyzed maximum 57.5 Hz with the stops removed.
    Therefore, the proposed TS changes do not significantly increase 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated. 
The proposed TS changes do not make any physical changes to the fuel, 
or the way the fuel responds to a transient or accident. The 
radiological barriers are not compromised. The fuel will continue to be 
operated to analyzed operating limits. No new failure mode is 
introduced.
    The proposed TS changes do not create new operating conditions that 
have not been evaluated. Removal of the Recirculation Master Flow 
Controller eliminates the possibility of a single failure initiated 
common mode event. Since the possibility of a common failure has been 
eliminated, the most limiting recirculation runout event is a one pump 
slow flow runout. This is the same kind of postulated accident as that 
previously evaluated, only it involves one pump instead of both pumps. 
Therefore, the proposed TS changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.

[[Page 48865]]

    3. The proposed TS changes do not involve a significant reduction 
in a margin of safety.
    The proposed TS changes do not make any physical changes to the 
fuel, or the way the fuel responds to a transient or accident. The 
radiological barriers are not compromised. The fuel will continue to be 
operated to analyzed operating limits. No new failure mode is 
introduced.
    Single pump runout based, generic flow biased ARTS thermal limits 
that do not take credit for MG set stops have been developed for LGS, 
Units 1 and 2. Adherence to approved ARTS-based flow biased thermal 
limits identified in the LGS, Units 1 and 2, COLRs and implemented in 
the plant process computer are sufficient to maintain the margin of 
safety as delineated in TS Sections 3/4.2.1, 3/4.2.3, and 3/4.2.4.
    Therefore, these proposed TS changes do not involve a significant 
reduction in a margin of safety.
    Based on the above review, the NRC staff concludes that 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: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: January 29, 1998.
    Description of amendment request: The amendment would delete the 
requirements for a security plan from the 10 CFR Part 50 license and 
technical specifications after the spent nuclear fuel is transferred to 
a Part 72 licensed independent spent fuel storage installation (ISFSI). 
Security requirements for the ISFSI would be in accordance with 10 CFR 
Part 72, Subpart H.
    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 physical structures, systems and components of the Trojan 
Nuclear Plant and the operating procedures for their use are 
unaffected by the proposed change. The proposed elimination of the 
security requirements for the 10 CFR Part 50 license, is predicated 
on approval of the Trojan ISFSI Security Plan (PGE 1073) which will 
be coincident with issuance of a 10 CFR Part 72 license and upon 
completion of the transfer of all nuclear fuel from the spent fuel 
pool to the ISFSI. The planned 10 CFR 72 licensing controls for the 
ISFSI will provide adequate confidence that personnel and equipment 
can perform satisfactorily for normal operations of the ISFSI and 
respond adequately to abnormal events/accidents. The proposed Trojan 
ISFSI Security Plan (PGE 1073) will also provide confidence that 
security personnel and safeguards systems will perform 
satisfactorily to ensure adequate protection for the storage of 
spent nuclear fuel. Therefore, the proposed 10 CFR Part 50 amendment 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change is security related, and as such, has no 
direct impact on plant equipment or the procedures for operating 
plant equipment and, therefore, does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. Because the proposed ISFSI area will be segregated from 
the 10 CFR Part 50 licensed area, licensed security activities under 
the 10 CFR Part 50 license will no longer be necessary after all the 
nuclear fuel has been moved. The planned 10 CFR 72 licensing 
controls for the ISFSI area will provide adequate confidence that 
personnel and equipment can perform satisfactorily for normal 
operations of the ISFSI and respond adequately to normal events/
accidents. Moreover, the ISFSI will be physically separate from the 
Trojan Nuclear Plant structures and equipment. Therefore, the 
proposed 10 CFR Part 50 license amendment 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 assumptions for a fuel handling and other accidents are not 
affected by the proposed license amendment. Because the proposed 
ISFSI area (that will contain the nuclear fuel) will be segregated 
from the 10 CFR Part 50 licensed area, licensed security activities 
under the 10 CFR Part 50 license will no longer be necessary. The 
planned 10 CFR 72 licensing controls for the ISFSI area will provide 
adequate confidence that personnel and equipment can perform 
satisfactorily for normal operations of the ISFSI and respond 
adequately to abnormal events/accidents. Also, the ISFSI will be 
physically separate from the Trojan Nuclear Plant structures and 
equipment. Therefore, the proposed 10 CFR Part 50 license amendment 
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: Branford Price Millar Library, 
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151, 
Portland, Oregon 97207.
    Attorney for licensee: Leonard A. Girard, Esq., Portland General 
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
    NRC Section Chief: Michael T. Masnik.

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: August 19, 1999. The August 19, 1999, 
submittal supersedes the February 18, 1999, submittal in its entirety 
(64 FR 14284).
    Description of amendment request: The proposed amendment would 
revise the Virgil C. Summer Nuclear Station (VCSNS) Technical 
Specifications (TS) to incorporate the new Pressure/Temperature (P-T) 
Limits Curves consistent with the analysis results of reactor vessel 
specimen W. These figures are contained in Section 3/4.4.9 and are 
presented as Figures 3.4-2 and 3.4-3. These figures were developed 
using the methodology included in WCAP 14040-NP-A, ``Methodology Used 
to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup 
and Cooldown Limit Curves,'' as well as Code Case N-640, ``Alternative 
Reference Fracture Toughness for Development of P-T Limit Curves for 
Section XI, Division I.'' A reduced flange temperature requirement was 
included in the development of the curves, with justification provided 
in WCAP 15102, Revision 1, ``V. C. Summer Unit I Heatup and Cooldown 
Limit Curves for Normal Operation.'' Additionally, the Bases section 
for the Pressure/Temperature Limits would be revised to accurately 
reflect current industry standards and regulations. A significant 
portion of this Bases section would be deleted due to the information 
also being located in WCAP 15102, Revision 1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards

[[Page 48866]]

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 specimen W. This analysis 
was performed using NRC approved methodology as documented in WCAP 
14040-NP-A, utilizing the 1996 ASME Boiler and Pressure Vessel Code, 
Section XI, Appendix G requirements, along with ASME Code Case N-
640. These curves provide the limits for operation of the Reactor 
Coolant System during heat up, cool down, criticality, and 
hydrotesting. These curves are provided without instrument 
uncertainties included, however, the uncertainties are included in 
the curves provided in the operating procedures. 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 heatup and cooldown curves to reflect the shift 
in nil-ductility reference temperature of the reactor vessel as a 
result of neutron embrittlement, and alternate methodology utilized 
to generate the curves. 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 accepted industry methodology, for a projected 
reactor vessel neutron exposure of 32 EFPY [Effective Full Power 
Years].
    The new curves will serve as the basis for operating 
limitations, to provide margin against non-ductile fractures. The 
uncertainties introduced by instrumentation, forced flow and 
elevation differences are not reflected in the TS curves. These 
uncertainties will be factored into the curves presented in the 
operating procedures. 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 Section Chief: Richard L. Emch, Jr.

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

    Date of amendment requests: August 11, 1999 (PCN-488).
    Description of amendment requests: The proposed amendments would 
modify the Technical Specifications for the San Onofre Nuclear 
Generating Station (SONGS) Units 2 and 3 to revise Surveillance 
Requirement (SR) 3.3.7.3 by providing allowable values in place of 
analytical limits for certain degraded voltage parameters, and by 
deleting unnecessary parameter limits in cases where plant safety is 
not affected. The proposed change would also delete redundant SR 
3.3.7.4.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No.
    Proposed Change Number (PCN)-488 revises the Technical 
Specification (TS) Surveillance Requirement (SR) acceptance criteria 
of the Loss of Voltage Signal (LOVS), Degraded Grid Voltage with 
Safety Injection Actuation Signal (DGVSS), and Sustained Degraded 
Voltage Signal (SDVS) relay circuits. These circuits are not 
accident initiators.
    PCN-488 revises the TS SR acceptance requirements to make them 
more limiting than the present requirements. Because the revised 
acceptance criteria are more limiting than the present requirements, 
the consequences of accidents analyzed in the Updated Final Safety 
Analysis Report (UFSAR) are not increased. PCN-488 also revises the 
TS SR acceptance requirements to delete upper and lower bounds in 
cases where the deleted bound provides no safety benefit. Deleting 
bounds having no safety significance does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    PCN-488 deletes redundant SR 3.3.7.4, which is not in NUREG-
1432, Standard Technical Specifications, Combustion Engineering 
Plants. Deleting a redundant requirement does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Consequently, the proposed amendment does not result in an 
increase in the probability of accidents evaluated in the UFSAR.
    2. Does this amendment request create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No.
    PCN-488 revises the TS SR acceptance criteria of the LOVS, 
DGVSS, and SDVS relay circuits, which are not accident initiators, 
and deletes a redundant SR. PCN-488 does not introduce any revision 
in the hardware configuration of the protective circuitry for LOVS, 
DGVSS or SDVS. The measurement required by the deleted, redundant 
surveillance is required elsewhere in the TS. For these reasons, 
PCN-488 does not create the possibility of any new or different kind 
of accident from any previously evaluated. '
    3. Does this amendment request involve a significant reduction 
in a margin of safety?
    No.
    PCN-488 provides allowable values for the acceptance criteria 
for the TS SR for LOVS, DGVSS and SDVS. As such, the revised values 
are more limiting than the current values, which represent design 
limits. Therefore, PCN-488 does not involve a significant reduction 
in a margin of safety.
    PCN-488 also revises the TS SR acceptance requirements to delete 
upper and lower bounds in cases where the deleted bound provides no 
safety benefit. Deleting bounds having no safety significance does 
not involve a significant reduction in a margin of safety.
    PCN-488 additionally deletes a redundant SR. Because the deleted 
surveillance is required elsewhere in the TS, this action does not 
involve a significant reduction in a margin of safety.
    For these reasons, PCN-488 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, Irvine, California 92713.

[[Page 48867]]

    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: August 31, 1998, as supplemented by 
letters dated April 19 and August 18, 1999. The August 31, 1998, 
application was originally noticed in the Federal Register on October 
21, 1998 (63 FR 56260).
    Description of amendment request: The proposed amendments would 
revise Technical Specification 3/4.4.9.3 by revising the cold 
overpressure mitigation curve to accommodate the replacement steam 
generators and by adding two surveillances (for the centrifugal 
charging pumps and the emergency core cooling system accumulators) to 
ensure the operability of the cold overpressure mitigation system.
    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.
    Reanalysis of STP [South Texas Project, Units 1 and 2] COMS 
[cold overpressure mitigation system] transients to consider design 
characteristics of Delta-94 RSGs [replacement steam generators] has 
shown that maximum allowable PORV [power-operated relief valve] 
setpoints decrease slightly, and continue to provide design basis 
low temperature overpressure protection with Delta-94 steam 
generators. This change request incorporates the new COMS curves 
into Technical Specification 3.4.9.3 (Figure 3.4-4). Maximum 
allowable PORV setpoints decrease with Delta-94 steam generators, 
and are conservative compared to Model E steam generator curves. Use 
of the new curves with either Model E or Delta-94 steam generators 
conforms to the STP design basis.
    These changes are based on a reanalysis that accounts for Model 
Delta-94 design, a decision to make calculation[s] of COMS maximum 
allowable PORV setpoint consistent with current industry standards 
as represented by WCAP-14040, and addition of two surveillances to 
the Technical Specification to ensure operability of COMS. Moving 
maximum allowable PORV setpoints in the conservative direction and 
adding surveillances to reinforce standard operating practice have 
no adverse effect on the probability or consequences of an accident 
previously evaluated. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed PORV maximum allowable setpoint changes do not 
create any new operating conditions or modes, and the added 
surveillances have no effect except to ensure operation of COMS as 
designed. The slight change to the maximum allowable PORV setpoint 
curves for the Cold Overpressure Mitigation System accommodates 
Delta-94 steam generator design characteristics, and COMS continues 
to perform in accordance with existing requirements, which are 
sufficient to ensure plant safety is preserved.
    The proposed change is the result of a reanalysis of a 
previously evaluated accident. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change reflects design characteristics of the new 
Delta-94 steam generators. The change to the COMS curves is in the 
conservative direction and does not affect any design failure point 
or system limitation. 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 
77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: August 18, 1999.
    Description of amendment request: The licensee proposed changing 
the Vermont Yankee Nuclear Power Station (VY) Technical Specifications 
by revising the reactor core spiral reloading pattern such that it 
begins around a source range monitor rather than from the center of the 
core. The offloading pattern would be the reverse sequence.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    VY has determined that the proposed change to reload the reactor 
core in a spiral pattern beginning around a Source Range Monitor 
(SRM) does not involve a significant increase in the probability or 
consequences of an accident previously evaluated. The design basis 
accident associated with refueling is the Refueling Accident; i.e., 
the accidental dropping of a fuel bundle onto the top of the core. 
There is no assumption as to the core loading pattern in the 
analysis of this accident. The analyzed abnormal operational 
transients associated with refueling are: (1) the Control Rod 
Removal Error During Refueling, and (2) the Fuel Assembly Insertion 
Error During Refueling. There is no assumption as to the core 
loading pattern in the analyses of these transients. The Fuel 
Assembly Insertion Error During Refueling transient involves 
mislocated and rotated fuel assembly loading errors. However, a 
change in the approved core loading pattern has no impact on the 
probability of mislocating or rotating a bundle while following that 
pattern. Furthermore, the proposed change implements a core loading 
pattern that provides improved flux monitoring as compared to the 
pattern prescribed by the current Technical Specifications. When 
loading the core in accordance with the proposed change, the SRM 
indication will be indicative of the true flux of the loaded fuel, 
as the creation of flux traps (moderator filled cavities surrounded 
on all sides by fuel) is precluded.
    The SRMs and the core loading pattern are not initiators of any 
accident previously evaluated. As such, the subject changes cannot 
affect the probability of an accident previously evaluated. The core 
loading pattern is not assumed in the mitigation of any accident. 
Since the proposed change provides improved flux monitoring by the 
SRMs, operators will have more accurate indication and SRM automatic 
trip functions will actuate based on a more accurate indication of 
flux. As such, any event mitigation function provided by the SRMs is 
enhanced by this change. Therefore, the associated changes do not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    VY has determined that the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. VY proposes to change the core reloading and 
offloading patterns to start and stop, respectively, at an

[[Page 48868]]

SRM versus the geometric center of the core as prescribed by current 
Technical Specifications. This ensures that flux monitoring 
instrumentation is always OPERABLE in the fueled region of the 
vessel. There is no separation of the monitoring device from the 
fuel by cavities of water as is the case with the pattern prescribed 
by the current Technical Specifications. As such, flux monitoring is 
enhanced during core reloading and offloading. This change is 
conservative relative to the current requirements. Therefore, no new 
or different kinds of accidents are created.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    VY has determined that the proposed change does not involve a 
significant reduction in a margin of safety. Loading around the 
geometric center of the core as prescribed by the current Technical 
Specifications results in cells of moderator separating the fuel 
from the instrumentation monitoring its flux. This change requires 
the flux monitoring instrumentation to be in the fueled region, and, 
in so doing, provides for more accurate monitoring of core flux 
during core reloading and offloading. As such, the operators will 
have more accurate indication and SRM automatic trip functions will 
actuate when the actual flux reaches the trip setpoints. Therefore, 
this change will not result in a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: August 18, 1999.
    Description of amendment request: The licensee proposed changing 
the Vermont Yankee Nuclear Power Station (VY) technical specifications 
(TSs) by revising the definition of the ``Surveillance Frequency'' to 
incorporate provisions that apply upon the discovery of a missed TS 
surveillance. The provisions would allow 24 hours to perform the 
surveillance before the applicable limiting condition for operation is 
entered.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    This change does not result in any physical alteration of plant 
systems, structures or components; nor does the change modify the 
manner in which plant equipment will be operated or maintained. As a 
result, the proposed change does not affect any of the parameters or 
conditions that contribute to the initiation or mitigation of any 
accidents previously evaluated.
    Surveillance frequencies are not assumed in the initiation of 
any analyzed event. Thus, conditions assumed in the plant accident 
analyses are unchanged. Furthermore, there is no relaxation of 
required setpoints or operating parameters.
    Therefore, the probability or consequences of an accident 
previously evaluated are not significantly increased since the most 
likely outcome of performing a surveillance is that it does, in 
fact, demonstrate the system or component is operable. VY has, 
therefore, determined that the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. The proposed change will not modify the 
physical plant or the modes of plant operation. The changes do not 
involve the addition or modification of equipment nor do they alter 
the design or operation of plant systems. These changes to Technical 
Specifications do not create any new or different kind of accident 
since they do not involve any change to the plant or the manner in 
which it is operated.
    Therefore, VY has determined that the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously [evaluated].
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    The proposed change does not affect design margins or 
assumptions used in accident analyses. The capability of safety 
systems to function and limiting safety system settings are 
similarly unaffected as a result of this change.
    The increased time allowed (up to 24 hours) for the performance 
of a surveillance discovered to have not been performed, is 
acceptable based on the small probability of an event requiring the 
associated component. The requested allowance will provide 
sufficient time to perform the missed surveillance in an orderly 
manner. Without the 24 hour delay, it is possible that the missed 
surveillance would force a plant shutdown; thus, the plant could be 
shutting down while the missed surveillance is being performed. As a 
result of this delay, the potential for human error will be reduced. 
Consequently, there is no significant reduction in a margin of 
safety as overall plant safety is enhanced due to the avoidance of 
unnecessary plant shutdowns.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Unit Nos. 1 and 2, Louisa County, Virginia

    Date of amendment request: August 4, 1999.
    Description of amendment request: The proposed changes to North 
Anna Power Station (NAPS) Units 1 and 2 Technical Specification (TS) 
4.4.1.6.1 and associated Bases will extend the drained reactor coolant 
loop verification time (verified as drained) from two hours to four 
hours prior to backfilling when returning the drained loop to service.
    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.
    Administrative procedures ensure that the initiation of seal 
injection in order to establish a partial vacuum in an isolated and 
drained loop will not create the potential for an inadvertent and 
undetected introduction of under-borated water into an isolated loop 
prior to returning the isolated loop to service. Additionally, 
extension of the drained loop verification time from two hours to 
four hours prior to backfill operations will not significantly 
diminish confidence that the isolated and drained loop will, in 
fact, be drained at the time the back-fill evolution is initiated. 
Therefore, there is no measurable increase in the probability or 
consequences of any accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated.

[[Page 48869]]

    There are no modifications to the plant as a result of the 
changes. No new accident or event initiators are created by the 
initiation of seal injection in order to establish a partial vacuum 
in an isolated and drained loop, and by the extension of the drained 
loop verification time requirement from two hours to four hours 
prior to backfill operations. Therefore, the proposed changes do not 
create the possibility of any accident or malfunction of a different 
type previously evaluated.
    3. Does the change involve a significant reduction in the margin 
of safety.
    The proposed changes have no effect on the safety analyses 
assumptions. Changes acknowledge the establishment of seal injection 
for the Reactor Coolant Pump in the isolated and drained loop as a 
prerequisite for the vacuum-assisted back-fill technique and extends 
the drained-loop verification time from two hours to four hours 
prior to backfill operations. The two hour interval was established 
to ensure that the drained loop is verified to be drained at a point 
in time sufficiently close to the initiation of the back-fill 
evolution such that no intervening event could occur that would 
render the loop no longer drained. Relaxation of the drained loop 
verification time from two hours to four hours will not 
significantly diminish confidence that the isolated and drained loop 
will be drained at the time the back-fill evolution is initiated. 
Therefore, the proposed changes do not result in a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams, 
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
23219.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: April 28, 1999.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) Section 3.4.A.4 and Table 4.1-
2B for Units 1 and 2. The proposed changes would reduce the minimum 
volume requirement for the refueling water chemical addition tank (CAT) 
to provide additional operating margin, and also correct administrative 
format errors in Table 4.1-2B.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Criterion 1--Does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The probability or the consequences of an accident previously 
evaluated are not increased. When the revised Safety Analysis Limit 
minimum CAT volume of 3800 gallons was implemented, consideration 
was given to the effects of the proposed reduced CAT volume on 
containment integrity analyses, containment spray and post-LOCA sump 
pH analyses, and the post-LOCA recirculation switchover time 
interval specified in Emergency Operating Procedures. The change was 
determined to be acceptable as accident analyses assumptions would 
continue to be met. The proposed TS minimum CAT volume (3930 
gallons) includes an allowance for the CAT level Channel Statistical 
Allowance (CSA), so that the safety analysis limit CAT volume (3800 
gallons) will not be violated when the measured CAT volume (i.e., 
tank level) is at or above the TS minimum CAT volume limit. The 
proposed reduction in the TS minimum CAT volume has no bearing on 
the probability of occurrence of any accident previously evaluated, 
since neither the volume nor the sodium hydroxide inventory of the 
CAT have any bearing on postulated accident initiators. Furthermore, 
because the affected accident analyses have been evaluated and found 
to meet their acceptance criteria with the reduced safety analysis 
limit CAT volume, the consequences of an accident previously 
evaluated is not increased.
    Criterion 2--Does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The possibility of a new or different kind of accident than any 
accident previously evaluated is not created. The proposed reduction 
in the TS minimum CAT volume does not involve any alterations to the 
physical plant that would introduce any new or unique operational 
modes or accident precursors. Only the TS minimum CAT volume is 
being changed to establish an operationally feasible alarm setpoint 
to provide the operators additional flexibility in maintaining the 
required CAT volume.
    Criterion 3--Does not involve a significant reduction in a 
margin of safety.
    The margin of safety is not reduced. It was determined that the 
affected safety analyses continue to meet their respective 
acceptance criteria with the revised minimum CAT volume. By 
implementing the proposed change in the TS minimum CAT volume, a CAT 
level alarm setpoint may be established which includes a 
conservative allowance for level measurement uncertainty such that 
neither the proposed TS minimum CAT volume nor the Safety Analysis 
Limit CAT volume will be violated at the time a CAT level alarm is 
received. Therefore, it is concluded that the proposed change will 
not reduce the margin of safety.
    This analysis demonstrates that the proposed amendment to the 
Surry Units 1 and 2 Technical Specifications does not involve a 
significant increase in the probability or consequences of a 
previously evaluated accident, does not create the possibility of a 
new or different kind of accident and 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 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: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams, 
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
23219.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: December 29, 1998, as supplemented by 
letter dated July 29, 1999. The December 29, 1998, amendment 
application was previously noticed in the Federal Register on February 
24, 1999 (64 FR 9023).
    Description of amendment request: The amendment would revise 
Section 5.6.6, ``Reactor Coolant System (RCS) Pressure and Temperature 
Limits Report (PTLR),'' of the improved Technical Specifications (TSs), 
that were issued in Amendment 123 on March 31, 1999. The amendment 
would (1) add the phrase ``and Cold Overpressure Mitigation System'' to 
the first sentence of item 5.6.6.b that identifies the limits that can 
be determined by the licensee in the PTLR, and (2) replace the current 
list of documents listed in item 5.6.6.b by the NRC letter that will 
approve this amendment and the Westinghouse report, WCAP-14040-NP-A, 
``Methodology Used to Develop Cold Overpressure Mitigation System 
Setpoints and RCS Heatup and Cooldown Limit Curves,'' dated January 
1996. WCAP-14040-NP-A is the NRC-approved topical report that provides 
a methodology for developing the cold overpressure mitigation system 
(COMS) setpoints and RCS heatup and cooldown limit curves for 
Westinghouse plants,

[[Page 48870]]

such as Wolf Creek Generating Station (WCGS).
    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.
    Incorporating the revised heatup and cooldown pressure/
temperature limit curves and the COMS PORV setpoint limit curve into 
the WCGS Technical Specifications does not affect the probability or 
consequences of an accident previously evaluated.
    The revised limit curves are calculated using the most limiting 
RTNDT for the reactor vessel components and include a 
radiation-induced shift corresponding to the end of the period for 
which the curves are generated. The COMS PORV Setpoint Limit Curve 
is calculated using the most limiting mass injection transient, 
taking into account operation of the NCP [normal charging pump] 
during shutdown modes. The changes do not affect the basis, 
initiating events, chronology, or availability/operability of safety 
related equipment required to mitigate transients and accidents 
analyzed for WCGS.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Adopting the revised limit curves redefines the range of 
acceptable operation for the Reactor Coolant System. This 
redefinition is a result of the analysis of reactor vessel 
surveillance specimens removed from the reactor in a continuing 
surveillance program which monitors the effects of neutron 
irradiation on the WCGS reactor vessel materials under actual 
operating conditions. Included in the revised limit curves is 
consideration for NCP operation during shutdown modes. Incorporating 
these revised curves does not create the possibility of an accident 
of a different type from any previously evaluated for WCGS.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The revision of these limit curves continues to maintain the 
margin of safety required for prevention of non-ductile failure of 
the WCGS reactor vessel during low temperature operation as required 
by 10 CFR 50, Appendices G and H. The revised curves primarily 
affect RCS operation below 350 deg.F by limiting the available 
pressure/temperature window for heatup and cooldown. The revised 
limit curves compensate for the in-service radiation induced 
embrittlement of the reactor vessel and accounts for the requirement 
that the closure flange region temperature must exceed the nil-
ductility temperature by at least 120 deg.F when pressure exceeds 
20% of the preservice hydrostatic test pressure.
    The revised COMS PORV Setpoint Limit Curve, which includes 
consideration of NCP operation during shutdown modes, ensures 
overpressure protection of the RCS and reactor vessel.

    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 locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Section Chief: Stephen Dembek.

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

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

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

    Date of amendment request: August 6, 1999.
    Brief description of amendment request: The proposed amendments 
would revise the Technical Specifications (TSs) contained in Appendix A 
to the Operating Licenses to incorporate a note into the TSs which will 
permit a one-time exemption, until September 30, 1999, from the 
90 deg.F limit stated in Surveillance Requirement (SR) 3.7.2.2. This SR 
currently requires that the average water temperature of the normal 
heat sink be less than or equal to 90 deg.F as demonstrated on a 24-
hour frequency. As stated in the proposed TS note, during the time 
period between approval and September 30, 1999, the average water 
temperature of the normal heat sink will be limited to less than or 
equal to 92 deg.F.
    Date of publication of individual notice in Federal Register: 
August 13, 1999 (64 FR 44243).
    Expiration date of individual notice: 14 days for comments, August 
27, 1999; 30 days for hearing, September 13, 1999.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (Regional Depository) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.

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 48871]]

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

    Date of application for amendment: July 30, 1999.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.7.8, ``Ultimate Heat Sink (UHS),'' to permit a 72-
hour delay in the UHS temperature restoration period prior to entering 
the plant shutdown required actions. This TS amendment is given as a 
temporary amendment change effective until September 30, 1999, after 
which the TS will revert back to the original TS provisions.
    Date of issuance: August 24, 1999.
    Effective date: August 24, 1999.
    Amendment No.: 184.
    Facility Operating License No. DPR-23: Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes (64 FR 43406 dated August 10, 1999). The 
notice provided an opportunity to submit comments on the Commission's 
proposed NSHC determination. No comments have been received. The notice 
also provided for an opportunity to request a hearing by September 8, 
1999, but indicated that if the Commission makes a final NSHC 
determination, any such hearing would take place after issuance of the 
amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and final determination of NSHC are contained in 
a Safety Evaluation dated August 24, 1999.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Sheri R. Peterson.

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

    Date of application for amendments: March 25, 1999.
    Brief description of amendments: The amendments revise various 
parts of the Technical Specifications (Appendix A of the Catawba 
operating licenses) to identify that the Trip Setpoints for the reactor 
trip system and engineered safety feature actuation system 
instrumentation are in reality Nominal Trip Setpoints.
    Date of issuance: August 13, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days from the date of issuance.
    Amendment Nos.: 179--Unit 1; 171--Unit 2.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24195).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 13, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power 
Station, Unit 2, Shippingport, Pennsylvania

    Date of application for amendment: June 18, 1996, as supplemented 
December 12, 1997, February 23, June 15, and July 15, 1999; and by 
separate application dated October 22, 1997, as supplemented February 
23, June 28, and July 15, 1999.
    Brief description of amendment: This amendment implements: (1) 
voltage-based repair criteria for BVPS-2 steam generator tubes similar 
to the changes approved for BVPS-1 in License Amendment No. 198. The 
changes revise BVPS-2 technical specifications (TSs) 4.4.5 and 3.4.6.2 
and associated Bases to reflect the guidance provided in the Nuclear 
Regulatory Commission's (NRC) Generic Letter 95-05, ``Voltage-Based 
Repair Criteria for Westinghouse Steam Generator Tubes Affected by 
Outside Diameter Stress Corrosion Cracking,'' (GL 95-05). Additionally, 
BVPS-2 TS Table 4.4-2 is revised to reference TS 6.6 for reporting 
requirements. (2) reduced reactor coolant system (RCS) specific 
activity limits in accordance with the NRC's guidance provided in GL 
95-05. The definition of Dose Equivalent I-131 is replaced with the 
Improved Standard TS definition in the first sentence, and an equation 
is added based on dose conversion derived from the International 
Commission on Radiation Protection (ICRP) ICRP-30. TS 3.4.8, Specific 
Activity, is revised by reducing the Dose Equivalent I-131 limit from 
1.0 [micro] Ci [curies]/gram to 0.35 [micro] Ci [curies]/gram for the 
48-hour limit and from 60 [micro] Ci [curies]/gram to 21 [micro] Ci 
[curies]/gram for the maximum instantaneous limit. Item 4.a in TS Table 
4.4-12, Primary Coolant Specific Activity Sample and Analysis Program; 
TS Figure 3.4-1, and the Bases for TS 3/4.4.8 are also modified to 
reflect the reduced Dose Equivalent I-131 limit.
    The February 23, 1999, letter provided a revised control room dose 
calculation in support of both the June 18, 1996, and October 22, 1997, 
amendment requests. Importantly, this calculation assumed the lower 
allowable primary-to-secondary leak rate limit associated with the June 
18, 1996, submittal, and the reduced RCS specific activity limits 
associated with the October 22, 1997, submittal. Because of this 
interdependence, the changes of the first amendment request must be 
implemented concurrently with those of the second in order for the 
supporting analysis to remain valid. Hence, both of these license 
amendment requests have been combined into this single amendment.
    Date of issuance: August 18, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No: 101.
    Facility Operating License No. NPF-73. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 64109) and March 25, 1998 (63 FR 14485). The December 12, 1997, 
February 23, June 15, June 28, and July 15, 1999, letters provided 
additional information but did not change the initial proposed no 
significant hazards consideration determinations or expand the 
amendment requests beyond the scope of the Federal Register notices.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 18, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

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

    Date of amendment request: November 24, 1998, as supplemented by 
letters dated February 25 and July 14, 1999.
    Brief description of amendments: The amendments revise the 
administrative sections of the Technical Specifications to reflect the 
approved consolidated quality assurance program, clarify the 
responsibilities of the shift technical advisor position on shift, 
simplify the contents of the monthly operating report description, 
complete the relocation of the fire protection requirements from

[[Page 48872]]

the Technical Specifications, and replace selected position titles with 
descriptions of functional responsibility.
    Date of issuance: August 26, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 198 and 209.
    Facility Operating License Nos. DPR-51 and NPF-6: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 27, 1999 (64 FR 
4156).
    The February 25 and July 14, 1999, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant, 
Unit No. 1, St. Lucie County, Florida

    Date of application for amendment: November 22, 1998.
    Brief description of amendment: This amendment revises the reactor 
thermal margin safety limit lines and flow rates stated in the St. 
Lucie, Unit 1, technical specifications (TS). The amendment also 
updates the reference for dose conversion factors used in Dose 
Equivalent Iodine-131 calculations, makes administrative changes to the 
criticality analysis uncertainty described in TS 5.6.1.a.1, updates the 
analytical methods used in determining core operating limits listed in 
TS 6.9.1.11, and revises the TS Bases for the steam generator pressure-
low trip setpoint.
    Date of Issuance: August 18, 1999.
    Effective Date: August 18, 1999.
    Amendment No.: 163.
    Facility Operating License No. NPF-16: Amendment revised the TS.
    Date of initial notice in Federal Register: February 10, 1999 (64 
FR 6696).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 18, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

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

    Date of application for amendment: October 30, 1998, as 
supplemented December 31, 1998, and May 12, 1999.
    Brief description of amendment: The amendment approves changes to 
the Improved Technical Specifications to reflect the use of Topical 
Report BAW-2421 for fluence determination and changes to the low 
temperature over-pressure protection limits. Changes to the CR-3 
Pressure/Temperature Limits Report to reflect plant operation to 32 
Effective Full Power Years were included in the submittal.
    Date of issuance: August 12, 1999.
    Effective date: As of date of issuance, to be implemented prior to 
commencing Cycle 12 operation.
    Amendment No.: 183.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 30, 1998 (63 
FR 71965). The supplemental letters dated December 31, 1998, and May 
12, 1999, did not change the original proposed no significant hazards 
consideration determination, or expand the scope of the amendment 
request as originally noticed.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 12, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.

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

    Date of application for amendment: November 30, 1998.
    Brief description of amendment: The Amendment revises Technical 
Specifications (TS) to allow both doors of the containment personnel 
air lock to be open during fuel movement and adds a provision for an 
outage equipment hatch.
    Date of issuance: August 16, 1999.
    Effective date: August 16, 1999.
    Amendment No.: 184.
    Facility Operating License No. DPR-31: Amendment revised the TS.
    Date of initial notice in Federal Register: January 27, 1999 (64 FR 
4157).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 16, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal River, Florida 34428.

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

    Date of application for amendment: July 30, 1998, as supplemented 
April 8 and July 8, 1999.
    Brief description of amendment: Revises Technical Specifications 
for the Control Room Emergency Ventilation System and the Ventilation 
Filter Test Program.
    Date of issuance: August 23, 1999.
    Effective date: August 23, 1999.
    Amendment No.: 185.
    Facility Operating License No. DPR-31: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 64115). The April 8 and July 8, 1999, supplements did not change the 
original proposed no significant hazards determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 23, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal River, Florida 34428.

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

    Date of application for amendment: December 3, 1998, as 
supplemented by letters dated March 26, April 16, May 7, May 21, June 
4, June 15, and June 29, 1999.
    Brief description of amendment: The amendment revises the Technical 
Specification Figure 2.1-1 ``Core Protection Safety Limit,'' and Figure 
2.1-3 ``Core Protection Safety Bases'' to reflect a decrease in reactor 
coolant system flow resulting from a revised analysis to allow 
operation of the TMI-1 facility with an average of 20 percent of the 
steam generator tubes plugged, and no more than 25 percent plugged in 
either generator.
    Date of issuance: August 19, 1999.
    Effective date: As of the date of demonstration of a satisfactory 
emergency feedwater pump flow test, as described in the license 
amendment and documented by the licensee, to be

[[Page 48873]]

performed during the 13R refueling outage scheduled to begin September 
10, 1999, and shall be implemented within 30 days of that date.
    Amendment No.: 214.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 30, 1998 (63 
FR 71967). The supplements dated March 26, April 16, May 7, May 21, 
June 4, June 15, and June 29, 1999, are within the scope of the 
original notice and do not change the proposed no significant hazards 
consideration finding.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 19, 1999.
    No significant hazards consideration comments received: No.
    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.

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

    Date of application for amendment: February 2, 1999 as supplemented 
July 29, 1999.
    Brief description of amendment: The amendment expands the scope of 
systems and test requirements for post-accident reactor building sump 
recirculation engineered safeguards features systems and increases the 
maximum allowable leakage of TS 4.5.4 from 0.6 gallons per hour (gph) 
to 15.0 gph.
    Date of issuance: August 24, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 215.
    Facility Operating License No. DPR-50. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 24, 1999 (64 FR 
14283).
    The supplemental letter did not change the initial no significant 
hazards consideration determination or the Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 24, 1999.
    No significant hazards consideration comments received: No.
    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.

Northeast Nuclear Energy Company, et al., Docket Nos. 50-336 and 50-
423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London 
County, Connecticut

    Date of application for amendment: March 5, 1999.
    Brief description of amendment: The amendments relocate certain 
Technical Specifications (TSs) Section 6.0 administrative controls to 
the NRC-approved Northeast Utilities Quality Assurance Program (NUQAP) 
Topical Report. Specifically, Sections 6.2.3 (Unit 3 only), 6.5, 6.6 
(partial), 6.7 (partial), and 6.10. The amendments also delete parts of 
Section 6.6 and 6.7 because their requirements are duplicated in 
existing regulations or elsewhere in the TSs. In addition, the 
amendments modify the table of contents and other TS sections to 
incorporate the aforementioned changes (e.g., correct references).
    Date of issuance: August 13, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 239 and 173.
    Facility Operating License Nos. DPR-65 and NPF-49: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 7, 1999 (64 FR 
17027).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 13, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

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

    Date of application for amendments: February 12, 1999, as 
supplemented July 8, 1999. The July 8, 1999, letter provided clarifying 
information and did not change the original no significant hazards 
consideration determination.
    Brief description of amendments: Administrative changes to correct 
typographical and editorial errors in Technical Specifications 
introduced in previous amendments.
    Date of issuance: August 23, 1999.
    Effective date: This license amendment is effective as of its date 
of issuance. The amendment will be implemented within 30 days.
    Amendments Nos.: 228 and 231.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24200).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 23, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (Regional Depository) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.

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

    Date of application for amendments: November 20, 1998, as 
supplemented by letter dated June 25, 1998.
    Brief description of amendments: These amendments modified 
technical specification surveillance requirement, 3.8.1.4, to allow 
increases in the minimum fuel oil required to be stored in the day 
tanks for emergency diesel generators.
    Date of issuance: August 23, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 185 and 159.
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 27, 1999 (64 FR 
4159).
    The supplemental letter provided clarifying information and did not 
change the initial no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 23, 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.

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

    Date of application for amendments: December 19, 1997, as 
supplemented June 1, 1998, and May 13, 1999.

[[Page 48874]]

    Brief description of amendments: The amendments revise TS 3.4.9, 
Pressurizer, to reduce the allowable pressurizer water volume for 
pressurizer operability. The allowable water volume is also revised to 
a percent pressurizer level of 57 percent.
    Date of issuance: August 19, 1999.
    Effective date: August 19, 1999, to be implemented within 30 days 
of issuance.
    Amendment Nos.: Unit 2--155; Unit 3--146.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 25, 1998 (63 FR 
14488).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 19, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.

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

    Date of application for amendments: September 4, 1998, as 
supplemented December 8, 1998, and February 16, 1999 (PCN 493).
    Brief description of amendments: The amendments revise Technical 
Specification 3.4.10, Pressurizer Safety Valves, to increase the as-
found pressurizer safety valve setpoint tolerances.
    Date of issuance: August 19, 1999.
    Effective date: August 19, 1999, to be implemented within 30 days 
of issuance.
    Amendment Nos.: Unit 2--156; Unit 3--147.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 10, 1999 (64 
FR 6711). The licensee's letters dated December 8, 1998, and February 
16, 1999, provided clarifications and additional information that were 
within the scope of the original Federal Register notice and did not 
change the staff's initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 19, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.

Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph M. 
Farley Nuclear Plant, Unit 1, Houston County, Alabama.

    Date of amendment request: April 23, 1999, as supplemented by 
letters dated July 22, July 30 and August 12, 1999.
    Brief Description of amendment: The amendment adds an additional 
condition to the license which allows Southern Nuclear Operating 
Company to operate Unit 1 for Cycle 16 based on a risk-informed 
approach to evaluate steam generator tube structural integrity.
    Date of issuance: August 17, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 143.
    Facility Operating License No. NPF-2: Amendment revises the 
Facility Operating License to add a license condition.
    Date of initial notice in Federal Register: June 16, 1999 (64 FR 
32291).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 17, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.

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

    Date of amendment request: March 22, 1999, as supplemented July 15, 
1999.
    Brief description of amendments: The amendments revised Technical 
Specification 3/4.7.1.6, ``Atmospheric Steam Relief Valves,'' and added 
a new Technical Specification for atmospheric steam relief valve 
instrumentation, to ensure that the automatic feature of the steam 
generator power-operated relief valves (i.e., the atmospheric steam 
relief valves) remains operable during Modes 1 and 2.
    Date of issuance: August 19, 1999.
    Effective date: August 19, 1999, to be implemented within 30 days.
    Amendment Nos.: Unit 1--114; Unit 2--102.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 21, 1999 (64 FR 
19565).
    The July 15, 1999, supplement provided revised Technical 
Specification pages and clarifying information that was within the 
scope of the original Federal Register notice and did not change the 
staff's initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 19, 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, Texas 77488.

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

    Date of application for amendment: June 3, 1999 (TS 397).
    Brief description of amendment: The Amendments change the Technical 
Specifications (TS) by reducing the Allowable Value used for Reactor 
Vessel Water Level--Low, Level 3 for several instrument functions.
    Date of issuance: August 16, 1999.
    Effective date: August 16, 1999.
    Amendment Nos.: 260 and 219.
    Facility Operating License Nos. DPR-52 and DPR-68: Amendments 
revise the TS.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38037).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 16, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Athens Public Library, 405 E. 
South Street, Athens, Alabama 35611.

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

    Date of application for amendment: April 16, 1999, as supplemented 
June 9, 1999.
    Brief description of amendment: The amendment clarifies the 
inservice inspection requirements regarding the granting of relief from 
the American Society of Mechanical Engineers (ASME) Code requirements 
by the NRC. The amendment also made changes to reflect previous NRC 
approval of the use of ASME Code Case N-560.
    Date of Issuance: August 13, 1999.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.

[[Page 48875]]

    Amendment No.: 172.
    Facility Operating License No. DPR-28. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38037).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated August 13, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

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

    Date of application for amendment: June 24, 1999.
    Brief description of amendment: The amendment clarifies the basis 
for the reactor protection system bypass of the turbine stop valve 
closure and turbine control valve fast closure scram signals at low 
power. The amendment clarifies that the analytical basis for this 
bypass corresponds to a fraction of reactor rated thermal power and not 
other measures of power, for instance, turbine power.
    Date of Issuance: August 13, 1999.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 173.
    Facility Operating License No. DPR-28.: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38038).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated August 13, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

    Dated at Rockville, Maryland, this 1st day of September 1999.

    For the Nuclear Regulatory Commission.
Suzanne C. Black,
Deputy Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 99-23300 Filed 9-7-99; 8:45 am]
BILLING CODE 7590-01-P