[Federal Register Volume 64, Number 36 (Wednesday, February 24, 1999)]
[Notices]
[Pages 9183-9209]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-4391]



[[Page 9183]]

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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 January 30, 1999, through February 11, 1999. 
The last biweekly notice was published on February 10, 1999.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By March 26, 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.

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    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendment request: January 28, 1999.
    Description of amendment request: The H. B. Robinson, Unit No. 2, 
Technical Specifications (TSs) are proposed to be changed to replace 
and add analytical methodologies used to determine acceptable core 
designs and provide inputs to methodologies that develop the core 
operating limits 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, 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 in a methodology have been previously 
generically reviewed and approved for use by the NRC for determining 
core neutronics design and gadolinimum oxide thermal conductivity. 
Analyzed events are assumed to be initiated by the failure of plant 
structures, systems, or components. The fuel design parameters 
developed in accordance with the new methodologies are bounded by 
the limitations in the NRC acceptance in its safety evaluations of 
the new methodologies. The topical reports associated with the new 
methodologies demonstrate that the integrity of the fuel will be 
maintained during normal operations and that design requirements 
preclude fuel rods containing gadolinium oxide from being limiting 
in accident and related safety analyses. The proposed change does 
not have a detrimental impact on the integrity of any plant 
structure, system, or component. The proposed change will not alter 
the operation of any plant equipment, or otherwise increase its 
failure probability. Therefore, the probability of occurrence for a 
previously analyzed accident is not significantly increased.
    The consequences of a previously analyzed accident are dependent 
on the initial conditions assumed for the analysis, the behavior of 
the fuel during the analyzed accident, the availability and 
successful functioning of the equipment assumed to operate in 
response to the analyzed event, and the setpoints at which these 
actions are initiated. The proposed changes to methodology continues 
to meet applicable design and safety analyses acceptance criteria 
for neutronics design analysis and gadolinimum oxide thermal 
conductivity. The topical reports associated with the new 
methodologies demonstrate that the integrity of the fuel will be 
maintained as is assumed or is bounded initially in accident 
analyses and that design requirements preclude fuel rods containing 
gadolinimum oxide from being limiting in accident and related safety 
analyses. The proposed change does not affect the performance of any 
equipment used to mitigate the consequences of an analyzed accident. 
As a result, no analyses assumptions are violated and there are no 
adverse effects on the factors that contribute to offsite or onsite 
dose as the result of an accident. The proposed change does not 
affect setpoints that initiate protective or mitigative actions. The 
proposed change ensures that plant structures, systems, or 
components are maintained consistent with the safety analysis and 
licensing bases. Based on this evaluation, there is no significant 
increase in the consequences of a previously analyzed event.
    Therefore, the proposed change does not involve any 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 change does not involve any physical alteration of 
plant systems, structures, or components. The proposed changes in 
methodology continue to meet applicable criteria for neutronics 
design analysis and assure that design requirements preclude fuel 
rods containing gadolinimum oxide from being limiting. The proposed 
change does not involve a physical alteration of the plant other 
than allowing for fuel design in accordance with NRC approved 
methodologies. No new or different equipment is being installed. No 
installed equipment is being operated in a different manner. There 
is no alteration to the parameters within which the plant is 
normally operated or in the setpoints that initiate protective or 
mitigative actions. As a result no new failure modes are being 
introduced. There are no changes in the methods governing normal 
plant operation, nor are the methods utilized to respond to plant 
transients altered. Therefore, the proposed change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The margin of safety is established through the design of the 
plant structures, systems, and components, through the parameters 
within which the plant is operated, through the establishment of the 
setpoints for the actuation of equipment relied upon to respond to 
an event, and through margins contained within the safety analyses. 
The proposed change is to methodologies that continue to meet 
applicable criteria for neutronics design analysis and continues to 
assure that design requirements preclude fuel rods containing 
gadolinimum oxide from being limiting. The proposed change does not 
impact the condition or performance of structures, systems, 
setpoints, and components relied upon for accident mitigation. The 
proposed change does not significantly impact any safety analysis 
assumptions or results. Therefore, the proposed change does not 
result in a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

[[Page 9185]]

    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Project Director: Cecil B. Thomas.

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

    Date of amendment request: November 25, 1998.
    Description of amendment request: The proposed amendments would 
revise Improved Technical Specifications 3.8.4 and 3.8.9 to support on-
line replacement of the Braidwood 125 Volt DC AT&T batteries with new 
Charter Systems Inc. batteries.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    During the replacement of the existing batteries, a temporary 
battery bank will provide the same function as the AT&T batteries 
being removed. Even though this temporary battery will not be 
seismically mounted, due to its location in the Turbine Building, it 
is the safety related AT&T battery which was previously qualified 
and used to perform this function on Unit 1.
    While the temporary battery is being connected, the DC bus will 
be supplied by the existing crosstie with Unit 1. Similar crosstie 
conditions are allowed under the present Improved Technical 
Specifications.
    The DC system is normally supplied by the AC system through the 
ESF [Engineered Safety Feature] battery charger. The essential 
function of the DC system battery is to supply control power 
necessary to start and load the Diesel Generators. Once the Diesel 
Generators are on line, the DC system will be supplied via the 
battery charger. However, the ESF batteries have been sized for one 
hour to provide additional assurance that the critical DC loads are 
available in the event of a loss of a battery charger.
    During the 10 day Completion Time when the temporary battery and 
the ESF charger are supporting the bus, the ability of that DC 
Division to mitigate an event/accident is unchanged except for its 
ability to cope with a seismic event. However, the probability of a 
seismic event concurrent with the 10 day Completion Time is 
extremely small. During a seismic event, one DC division may be 
compromised, however, the unit has adequate DC power available in 
the form of the other division to mitigate all Design Basis 
accidents. This loss of one DC division is bounded by the loss of an 
entire AC division, a condition which the plant is currently 
evaluated to withstand.
    During the 8 hour Completion Time to connect and disconnect the 
temporary battery, there is no adverse impact on Unit 1. The 
compensatory measures to manually open the crosstie will ensure the 
Unit 1 DC battery can supply its required loads for the entire one 
hour duty cycle. The Unit 2 DC bus, which is crosstied, will be de-
energized in the event of a Unit 2 accident based on the 
compensatory measures. This action would only be required if the 
associated Diesel Generator were to fail to re-energize its 
associated charger. This condition is consistent with the other 
crosstie scenarios currently permitted by the Technical 
Specifications. Thus, the 8 hour Completion Time is consistent with 
the two hour Completion Time with respect to the ability to safely 
shutdown the Unit. Only the duration of the Completion Time is 
different.
    Based on the above, the overall design, function, and operation 
of the DC system and equipment has not been significantly modified 
by these changes. The proposed changes do not affect any accident 
initiators or precursors and do not alter the design assumptions for 
the systems or components used to mitigate the consequences of an 
accident as analyzed in UFSAR Chapter 15.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    During the replacement of the existing batteries, a temporary 
battery bank will provide the same function as the batteries being 
removed. Even though this temporary battery is not seismically 
mounted, it is the safety related AT&T battery which was previously 
qualified and utilized to perform this function on Unit 1. Because 
this temporary battery is identical to the battery that is currently 
installed, and will be connected and used in the same way, no new 
electrical or functional failure modes are created.
    The temporary battery will be located in the turbine building, 
which is non-seismic. The temporary battery will not be seismically 
mounted. Thus, a seismic failure of the batteries is possible. Since 
the temporary battery is located in the turbine building the 
potential for battery failure to initiate an accident is not 
present, and failure of the battery cannot create a different 
response from any previously postulated accident.
    Due to the location of the main generator in relationship to the 
temporary batteries, a turbine blade failure would not hit the 
battery unless it penetrated the turbine casing and ricocheted in 
the direction of the battery, which is an unlikely scenario due to 
the orientation of the temporary battery. Likewise, an unmitigated 
Outside Containment Steam Line Break of either unit would be 
interrupted by the successful closure of all MSIVs [Main Steam 
Isolation Valves] thereby leaving the battery and the DC bus intact 
and available. Also any affects of a postulated storm on the turbine 
building have been previously addressed and would not change as a 
result of the batteries being temporary located there.
    While the temporary battery is being connected, the DC bus will 
be supplied by the existing crosstie with Unit 1. To prevent any 
occurrence on Unit 2 from adversely affecting Unit 1, this crosstie 
will be manually disconnected based on specific criteria that may be 
indicative of a Unit 2 accident (specifically a Unit 2 LOOP). Once 
the crosstie is opened, the Unit 2 bus will be de-energized and the 
other Unit 2 division will be required to mitigate the accident. 
This loss of one DC division is bounded by the loss of one division 
(AC or DC), a condition which the plant is currently evaluated to 
withstand.
    The DC system and its equipment will continue to perform the 
same function and be operated in the same fashion. The proposed 
changes do not introduce any new accident initiators or precursors, 
or any new design assumptions for the systems or components used to 
mitigate the consequences of an accident. Therefore, the possibility 
of a new or different kind of accident from any accident previously 
evaluated has not been created.
    Therefore, this proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    During the replacement of the existing batteries, a temporary 
safety related battery bank will perform the same function as the 
batteries being removed. Even though this temporary battery is not 
seismically mounted, it is the safety related battery which was 
previously qualified and used to perform this function on Unit 1 and 
is identical to the safety related battery that is currently 
installed. Therefore, it has the same capacity, margin and 
capability to fulfill the requirements of the Unit 2 DC bus as the 
existing qualified battery. The proposed replacement activity will 
not prevent the plant from responding to either a seismic event or 
design basis accident. In both cases, the design mitigation 
capability will be maintained. Due to the limited duration of the 
activity and the planned contingency actions, a significant 
reduction in the margin of safety will not result.
    While the temporary battery is being connected, the DC bus will 
be supplied by the existing crosstie with Unit 1. This condition is 
currently allowed for a limited time by the Improved Technical 
Specifications.
    The inherent design conservatism of the DC system and its 
equipment has not been altered. The DC system and its equipment will 
continue to be operated with the same degree of conservatism. 
Accordingly, there is no significant reduction in the margin of 
safety.

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

[[Page 9186]]

satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Wilmington Public Library, 201 
S. Kankakee Street, Wilmington, Illinois 60481.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Project Director: Stuart A. Richards.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of amendment request: December 29, 1998.
    Description of amendment request: The proposed amendments would 
revise the Technical Specification Tables 3.3.1-1 and 3.3.2-1, to 
revise twelve Reactor Trip System and Engineered Safety Feature 
Actuation System Allowable Values.
    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?
    These changes to the twelve AVs [Allowable Values] do not 
involve an increase in the probability of an accident previously 
evaluated. The AVs provide the basis for determining instrument 
channel operability and do not change the system function, or 
channel operation or calibration. Operation within the AV ensures 
the instrument channel's ability to provide the required reactor 
trip or engineered safety feature actuation signal during plant 
operation. In all cases, the proposed changes only make the twelve 
AVs more restrictive with respect to the current AVs, and do not 
effect the response characteristics of the instrumentation because 
actual trip setpoints are unchanged. There is no change being made 
to the approved design, nor is there any operational change being 
made which would increase the probability of occurrence of an 
accident previously evaluated. The RTS [Reactor Trip System] and 
ESFAS [Engineered Safety Feature Actuation System] systems which are 
actuated by the corresponding instrumentation setpoints will operate 
in the same manner as before and within their design limits.
    These changes to the twelve AVs do not involve an increase in 
the consequences of an accident previously evaluated. These changes 
have no effect on plant operation. There is no physical or 
operational change being made which would alter the sequence of 
events, plant response, or assumptions or conclusions of the 
affected analyses. The use of the AVs as a basis for determining 
instrument or channel operability does not change system operation 
or channel function. The proposed changes do not change the 
established trip setpoints for these functions. No design analyses 
have changed or will be affected. The twelve revised AVs are more 
restrictive than the current AVs and continue to ensure that the 
safety limits are not violated during anticipated transients, and 
that the consequences of design basis accidents remain acceptable. 
The change to the AVs does not degrade or prevent any actions from 
taking place in response to an accident. The use of NRC approved or 
endorsed methodology in developing the proposed AVs ensures that the 
present analytical limits for all accidents will be maintained. 
These proposed changes to the AVs for RTS and ESFAS instrumentation 
will continue to ensure that the associated RTS trip or ESFAS 
actuation signals will be generated when required within the bounds 
of the plant safety analyses. There is no change in the type or 
amount of any effluents released, and no change in either the onsite 
or offsite dose consequences as a result of this change.
    Therefore, based on this evaluation, this proposed amendment 
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?
    These proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed changes to the twelve AVs for RTS and ESFAS 
instrumentation will not affect the trip setpoints at which a 
reactor trip or engineered safety feature actuation is initiated. 
The trip setpoints contained in the Technical Requirements Manual 
are not being changed and will continue to be maintained. The only 
changes being made are to the AVs used as a basis for determining 
instrument channel operability. Because the trip setpoints are 
unchanged, RTS or ESFAS setpoint actuation is not affected by the 
revised AVs.
    An RTS trip or ESFAS actuation signal that may initiate between 
its trip setpoint and the associated AV is acceptable because an 
allowance has been made in the affected instrument uncertainty 
calculation to accommodate this deviation. It allows for potential 
drift while ensuring plant operation in a safe manner. Using this 
methodology provides plant operational flexibility and yet remains 
within the allowances accounted for in the various accident 
analyses. No new equipment is being installed, and no installed 
equipment is being operated in a new or different manner with these 
twelve AV changes. The revised AVs do not alter the intended design 
or operation of systems or instrument channels.
    As no physical plant equipment changes are being made, no new 
equipment failure modes are being introduced as a result of these 
proposed changes. There is no change in plant operation that affects 
previously evaluated failure modes and no change in plant response 
to a transient condition. These changes do not represent a new 
failure mode over what has been previously evaluated.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    There is no significant reduction in the margin of safety from 
these proposed changes. These proposed changes move twelve AVs 
closer to the trip setpoints compared to the existing AVs, which 
increases the margin of safety. An RTS trip or ESFAS actuation 
signal that may initiate between its trip setpoint and the 
associated AV is acceptable because an allowance has been made in 
the affected instrument uncertainty calculation to accommodate this 
deviation. The revised AVs have been calculated using NRC approved 
or endorsed methodology, which is consistent with existing safety 
analyses that define the margin of safety. Safety analyses 
assumptions and results are not affected.
    Therefore, these changes do not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Project Director: Stuart A. Richards.

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

    Date of amendment request: January 21, 1999.
    Description of amendment request: This amendment request proposes 
to relocate Technical Specification (TS) Section 3/4.6.I to the Updated 
Final Safety Analysis Report (UFSAR) and plant procedures. TS Section 
3/4.6.I contains reactor coolant chemistry limiting conditions for 
operation (LCO) and surveillance requirements (SR) for conductivity, 
chloride concentration and pH.

[[Page 9187]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Project Director: Stuart A. Richards.

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

    Date of amendment request: January 28, 1999.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) to correct Surveillance 
Requirement (SR) 3.7.13.4 and the associated Bases. This SR currently 
is incorrect and does not reflect the Fuel Handling Ventilation Exhaust 
System (FHVES) as designed. Specifically, the FHVES flow rate 
requirement has been inadvertently stated at half the design value 
(18,221 instead of 36,443 cfm [cubic feet per minute]). The proposed 
amendments would only revise the SR to the correct design value; no 
physical change to the FHVES design is involved.
    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. The FHVES is not an 
accident initiating system; therefore, there will be no impact on 
any accident probabilities by the approval of this amendment. The 
design of the system is not being modified by this proposed 
amendment. The amendment merely aligns TS requirements with the 
existing design and function of the system. Therefore, there will be 
no impact on any accident 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 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 FHVES is already capable of performing as 
designed. No safety margins will be impacted.
    Based upon the preceding analysis, Duke Energy has concluded 
that the proposed amendment does not involve a significant hazards 
consideration.

    The staff reviewed the licensee's analysis, and agrees 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: Mr. Paul R. Newton, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: December 16, 1998, supplemented January 
25, 1999.
    Description of amendment request: The proposed amendments would 
completely replace the High Pressure Injection (HPI) section of the 
Improved Technical Specifications that were issued on December 16, 
1998. The proposed changes would: (1) expand the applicability for the 
requirements regarding the third HPI pump, discharge crossover valves, 
and the HPI suction headers; (2) specify the HPI conditions and allowed 
times that require the discharge headers be cross-connected or 
separated; (3) incorporate limiting conditions for operation when 
specified equipment was inoperable during specified plant conditions; 
(4) specify changes in HPI system discharge path valve lineup when 
certain equipment is inoperable; (5) change the requirement to reduce 
reactor power when an HPI system is inoperable from 60 percent power to 
75 percent power and specify the length of time operation may continue 
at this power level; (6) address the failure to cross-connect the HPI

[[Page 9188]]

discharge headers as an independent condition; (7) add a requirement to 
verify by administrative means that the Atmospheric Dump Valve flow 
path for each steam generator is operable every 12 hours under certain 
conditions; (8) add a requirement that the HPI pump and crossover 
valves be restored to operable status within 30 days; (9) delete the 
requirement to restore the capability to automatically actuate the HPI 
within 24 hours; (10) add a Required Action to reduce reactor power to 
less than or equal to 75 percent power within 3 hours in the event an 
HPI train cannot be actuated by automatic or manual means; (11) expand 
the Completion Time for restoring an inoperable HPI train to 72 hours; 
(12) require that Limiting Condition for Operation 3.0.3 be entered 
immediately if two HPI trains or two HPI (low pressure injection) -LPI 
flow paths are inoperable; (13) change the surveillance requirement to 
manually cycle open each LPI-HPI flow path discharge valve every 18 
months to require that the HPI discharge crossover valves be cycled 
every 18 months; and (14) add or modify various administrative and 
Bases changes that support the proposed changes. The licensee supplied 
data resulting from risk-informed analyses that were performed in 
accordance with Regulatory Guides 1.174 and 1.177 to support the 
evaluation.
    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:
    No. The proposed change do not involve a physical alteration of 
the plant. No new or different equipment is being installed, and no 
installed equipment is being operated in a new or different manner. 
No set points for parameters which initiate protective or mitigative 
action are being changed.
    The proposed changes do not have any impact upon the ability of 
the HPI [High Pressure Injection] System to add soluble poison to 
the Reactor Coolant System. The remaining potential impact is upon 
the ability to mitigate the consequences of a small break LOCA 
[Loss-of-Coolant Accident], which is addressed below. The small 
break LOCA is the limiting design basis accident with respect to HPI 
System operability requirements.
    The Technical Specification requirements for the HPI System are 
supported by a spectrum of small break LOCA analyses based on the 
approved Evaluation Model described in FTI [Framatome Technologies 
Incorporated] topical report BAW-10192PA. These small break LOCA 
analyses demonstrate that the acceptance criteria of 10 CFR 50.46 
are satisfied.
    The requirements of LCO [Limiting Condition for Operation] 3.5.2 
assure that flow can be provided via two HPI trains (i.e., one HPI 
train responds automatically upon an ESPS [Engineered Safeguards 
Protective System] signal, and the second HPI train is aligned 
within 10 minutes via operator actions in the Control Room) 
following a small break LOCA and a single active failure. The full 
power small break LOCA analyses supporting this proposed license 
amendment have been performed in accordance with the approved 
Evaluation Model described in FTI topical report BAW-10192P.
    If enhanced steam generator cooling is not credited in the 
accident analysis, two HPI trains are required to mitigate specific 
small break LOCAs with Thermal Power [less than or equal to] 75% RTP 
[Reactor Thermal Power]. However, if equipment not qualified as QA-1 
(i.e., an ADV [Atmosphic Dump Valve] flow path for one steam 
generator) is credited for enhanced steam generator cooling, the 
safety analyses have determined that the capacity of one HPI train 
is sufficient to mitigate a small break LOCA on the discharge of the 
reactor coolant pumps if Thermal Power [less than or equal to] 75% 
RTP. An ADV flow path for each steam generator is credited as a 
compensatory measure in Actions B and C of LCO 3.5.2 to permit 
operation to continue with THERMAL POWER [less than or equal to] 75% 
RTP: a) for 30 days with an HPI pump of one or more HPI discharge 
crossover valve(s) inoperable; and b) for 72 hours with one HPI 
train inoperable. This provides additional defense-in-depth, because 
the ADV flow path for each steam generator is required to be 
operable while only one is needed to perform the function. 
Additionally, a risk-informed assessment (provided as Attachment 7 
to Duke's license amendment request dated December 18, 1998) 
concluded that operating the plant in accordance with the Required 
Actions was acceptable.
    The proposed changes involve crediting an additional operator 
action (i.e., steaming that steam generator through an ADV flow 
path) that has not previously been reviewed and approved by the 
staff for licensing basis small break LOCA analyses. Additionally, 
while the EFW System has been credited in past SBLOCA [small break 
LOCA] analyses as described in responses to NUREG-0565, actions to 
raise steam generator levels to the loss of subcooled margin 
setpoint were only assumed in the smaller SBLOCAs. These operator 
actions have been included in the Emergency Operating Procedure 
(i.e., AP/1, 2, or 3/A/1800/001) for many years.
    The times for completing these operator actions (i.e., feeding a 
steam generator via EFW [Emergency Feedwater] and steaming that 
steam generator through an ADV flow path) are new to the small break 
LOCA analysis and the licensing basis, and are considered 
reasonable. Crediting the performance of these operator actions 
within the specified time frames in the SBLOCA analyses does not 
result in any substantive change to the operator's response to [an] 
SBLOCA.
    In summary, the technical analyses described in this license 
amendment justify the adequacy of this specification and assure that 
operability of the HPI System is maintained in a manner consistent 
with the requirements of the design basis accidents. Therefore, it 
is concluded that this amendment request will not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated:
    No. The proposed changes do not involve a physical alteration of 
the plant. No new or different equipment is being installed, and no 
installed equipment is being operated in a new or different manner. 
No set points for parameters which initiate protective or mitigative 
action are being changed. As a result, no new failure modes are 
being introduced.
    The requirements of ITS [Improved Technical Specification] 3.5.2 
continue to assure that operability of the HPI System is maintained 
in a manner consistent with the requirements of the design basis 
accidents. The requirements are supported by small break LOCA 
analyses which demonstrate that the acceptance criteria of 10 CFR 
50.46 are satisfied.
    The proposed change involve crediting an additional operator 
action (i.e., steaming that steam generator through an ADV flow 
path) that has not previously been reviewed and approved by the 
staff for licensing basis small break LOCA analyses. Additionally, 
while the EFW System has been credited in past SBLOCA analyses as 
described in responses to NUREG-0565, actions to raise steam 
generator levels to the loss of subcooled margin setpoint were only 
assumed in the smaller SBLOCAs. These operator actions have been 
included in the Emergency Operating Procedure (i.e., AP/1, 2, or 3/
A/1800/001) for many years.
    The times for completing these operator actions (i.e., feeding a 
steam generator via EFW and steaming that steam generator through an 
ADV flow path) are new to the small break LOCA analysis and the 
licensing basis, and are considered reasonable. Crediting the 
performance of these operator actions within the specified time 
frames in the SBLOCA analyses does not result in any substantive 
change to the operator's response to [an] SBLOCA.
    Therefore, this proposed amendment will not create the 
possibility of any new or different kind of accident.
    (3) Involve a significant reduction in a margin of safety.
    No. The requirements of ITS 3.5.2 continue to assure that 
operability of the HPI System is maintained in a manner consistent 
with the requirements of the design basis accidents. The 
requirements are supported by small break LOCA analyses which 
demonstrate that the acceptance criteria of 10 CFR 50.46 are 
satisfied. These analyses were performed in accordance with the 
Evaluation Model described in FTI topical report BAW-10192P.
    Therefore, it is concluded that the proposed amendment request 
will not result in a significant decrease in the margin of safety.


[[Page 9189]]


    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC.
    NRC Project Director: Herbert N. Berkow.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver 
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

    Date of amendment request: January 18, 1999
    Description of amendment request: The proposed amendments would: 
(1) delete license condition 2.C.(3) from the Beaver Valley Power 
Station, Unit No. 1 (BVPS-1) operating license and delete some 
references to two-loop operation from BVPS-1 Technical Specifications 
(TSs); (2) revise BVPS-1 and Beaver Valley Power Station, Unit No. 2 
(BVPS-2) TS 2.2.1, 3.3.2.1, associated tables 2.2-1 and 3.3.4, and 
associated bases, to use consistent format and wording between units; 
(3) revise BVPS-1 and BVPS-2 TS 2.2.1, 3.3.2.1, associated tables 2.2-1 
and 3.3.4, and associated bases, to include revised nominal trip 
setpoints and allowable values which are more conservative than those 
currently listed; (4) delete or revise TS to reflect the current 
configuration of Unit 1 plant hardware; and (5) make miscellaneous 
editorial changes to BVPS-1 and BVPS-2 TS and associated Bases to 
define terms, revise formatting, modify titles, and add license numbers 
to pages.
    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 [as modified by the NRC staff 
based upon information provided elsewhere in the licensee's submittal].

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    This proposed amendment includes changes to nominal Reactor Trip 
System (RTS) and Engineered Safety Feature Actuation System (ESFAS) 
trip setpoints and allowable values that have been determined with 
the use of an approved methodology. The new values ensure that all 
automatic protective actions will be initiated at or before the 
condition assumed in the safety analysis. This change, which 
includes modification of the requirements stated in Limiting Safety 
System Setting (LSSS) 2.2.1 and Limiting Condition for Operation 
(LCO) 3.3.2.1, will allow the nominal trip setpoints to be adjusted 
within the calibration tolerance band allowed by the setpoint 
methodology. There will be no adverse effect on the ability of the 
channels to perform their safety functions as assumed in the safety 
analyses. Since there will be no adverse effect on the trip 
setpoints or the instrumentation associated with the trip setpoints, 
there will be no significant increase in the probability of any 
accident previously evaluated.
    Other changes in trip system function, content and format are 
proposed based on the current configuration of the trip system 
hardware at Beaver Valley Power Station (BVPS) Unit No. 1. 
Similarly, since the ability of the instrumentation to perform its 
safety function is not adversely affected, there will be no 
significant increase in the consequences of any accident previously 
evaluated.
    Since the safety analysis is unaffected by this change there is 
no change in the consequences of any previously evaluated accident.
    The editorial changes do not affect plant safety. The 
administrative change, for BVPS Unit 1 only, pertaining to two loop 
operation and Reactor Coolant System isolation valve position, does 
not affect plant safety. The Technical Specification requirements in 
LCOs 3.4.1.1 and 3.4.1.4.1 will continue to [prohibit two-loop 
operation and] ensure safe plant operation by properly controlling 
the operation and position of the reactor coolant loops and Reactor 
Coolant System isolation valves.

[The administrative change to delete line item 7.d, pertaining to 
Auxiliary Feedwater (AFW) Pump Auto-start on Emergency Bus 
Undervoltage, from BVPS-1 TS Tables 3.3-3, 3.3-4, and 4.3-2 will not 
affect plant safety because this function is not directly initiated 
by bus undervoltage. Rather, the automatic start of the motor-driven 
AFW pumps is accomplished by the combination of 1) Emergency Bus 
feed breaker opening 2) valid start signal from ESFAS, and 3) 
Emergency Diesel Generator (EDG) sequencer actuation. Requirements 
for these items are included in the ESFAS related TS, Table 3.3-3 
and 3.3-4 items 7.a, 7.c, 7.e, and EDG related TS 4.8.1.1.2.b.3 (b). 
Therefore, since there is no change made to the plant hardware or 
its operation and requirements related to the AFW pump auto-start 
function are maintained elsewhere in the BVPS-1 TS, deleting line 
item 7.d from BVPS-1 TS Tables 3.3-3, 3.3-4, and 4.3-2 will not 
change the probability or consequences of any accident previously 
evaluated.]
    Therefore, this change does not involve any significant increase 
in the probability of occurrence of any accident previously 
evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed amendment includes changes to the format and 
magnitudes of nominal trip setpoints and allowable values that 
preserve all safety analysis assumptions related to accident 
mitigation. The protection system will continue to initiate the 
protective actions as assumed in the safety analysis. The proposed 
changes to LSSS 2.2.1 and LCO 3.3.2.1 will continue to ensure that 
the trip setpoints are maintained consistent with the setpoint 
methodology and the plant safety analysis. This proposed amendment 
does not involve additional hardware changes. Plant operation will 
not be changed.
    Other proposed changes are made so that the Technical 
Specifications more accurately reflect the plant-specific trip 
system hardware in BVPS Unit No. 1.
    Furthermore, the proposed changes do not alter the functioning 
of the RTS and ESFAS. Therefore, the proposed change does not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed RTS and ESFAS trip setpoints 
are calculated with an approved methodology. The proposed changes to 
LSSS 2.2.1 and LCO 3.3.2.1 will continue to ensure that the trip 
setpoints are maintained consistent with the setpoint methodology 
and the plant safety analysis. Therefore, the response of the RTS 
and ESFAS to accident transients reported in the Updated Final 
Safety Analysis Report is unaffected by this change. No additional 
hardware changes are involved. Therefore, accident analysis 
acceptance criteria are not affected. Other proposed changes are 
made so that the protection system Technical Specifications more 
accurately reflect the plant-specific trip system hardware in BVPS 
Unit No. 1.
    The editorial changes do not affect plant safety. The 
administrative change, for BVPS Unit 1 only, pertaining to two loop 
operation and Reactor Coolant System isolation valve position, does 
not affect plant safety. The Technical Specification requirements in 
LCOs 3.4.1.1 and 3.4.1.4.1 will continue to [prohibit two-loop 
operation and] ensure safe plant operation by properly controlling 
the operation and position of the reactor coolant loops and Reactor 
Coolant System isolation valve.

[The administrative change to delete line item 7.d, pertaining to 
Auxiliary Feedwater (AFW) Pump Auto-start on Emergency Bus 
Undervoltage, from BVPS-1 TS Tables 3.3-3, 3.3-4, and 4.3-2 will not 
affect plant safety because this function is not directly initiated 
by bus undervoltage. Rather, the automatic start of the motor-driven 
AFW pumps is accomplished by the combination of (1) Emergency Bus 
feed breaker opening, (2) valid start signal from ESFAS, and (3) EDG 
sequencer actuation. Requirements for these items are included in 
the ESFAS related TS, Table 3.3-3 and 3.3-4 items 7.a, 7.c, 7.e, and 
EDG related TS 4.8.1.1.2.b.3 (b). Therefore, since there is no 
change made to the plant hardware or its operation and requirements 
related to the AFW pump auto-start function are maintained elsewhere 
in the BVPS-1 TS,

[[Page 9190]]

deleting line item 7.d from BVPS-1 TS Tables 3.3-3, 3.3-4, and 4.3-2 
will not involve a significant reduction in a margin of safety.]
    Therefore, operation of the facility in accordance with the 
proposed amendment will not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: B.F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: S. Singh Bajwa.

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

    Date of amendment request: December 16, 1998.
    Description of amendment request: The licensee has proposed an 
amendment of Facility Operating License No. NPF-47, Appendix A--
Technical Specifications, Section 2.1.1.2, entitled ``Reactor Core 
[Safety Limits].'' The proposed amendment will change the two 
recirculation loop Minimum Critical Power Ratio (MCPR) limit from 1.13 
to 1.12 and the single recirculation loop MCPR limit from 1.14 to 1.13. 
The revised limits are necessary to address the operation of Cycle 9 
following the refueling outage which is scheduled to begin April 1999.
    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 request does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The plant/cycle specific SLMCPRs have been calculated using 
methods identical to those used by General Electric (GE) to assess 
the SLMCPR for other Boiling Water Reactors (BWRs). Similar methods 
were used to determine the value of the SLMCPR for the previous 
cycle. These methods are within the existing design and licensing 
basis and cannot increase the probability or severity of an 
accident. The basis of the SLMCPR calculation is to ensure that 
greater than 99.9% of all fuel rods in the core avoid transition 
boiling and fuel damage in the event of the occurrence of 
Anticipated Operational Occurrences (AOO) or a postulated accident.
    The SLMCPR is used to establish the Operating Limit Minimum 
Critical Power Ratio (OLMCPR). Neither the SLMCPR nor the OLMCPR are 
initiators or affect initiators of an accident previously evaluated 
and therefore changes to the SLMCPR do not increase the probability 
of any accident previously evaluated. The proposed changes involve 
the use of an accepted methodology in calculating the SLMCPR and, 
since there is no change in the definition of the SLMCPR, these 
changes will not affect the consequences of any accident previously 
evaluated. In addition, the proposed changes do not involve any 
change in the way the plant is operated. Existing procedures will 
ensure that the SLMCPR is not violated. Therefore, these changes 
have no effect on the consequences of an accident.
    On these bases, there will be no increase in the probability or 
consequences of an accident previously analyzed as a result the 
proposed changes.
    2. The request does not create the possibility of occurrence of 
a new or different kind of accident from any accident previously 
evaluated.
    The proposed changes consist of SLMCPR calculated from an 
accepted method of analysis that has been used by many BWRs. These 
changes do not involve any alteration of the plant and do not affect 
the plant operation. Neither the SLMCPR nor the OLMCPR can initiate 
an event, therefore a change to the SLMCPR does not create the 
possibility of occurrence of a new or different kind of accident 
from any accident previously evaluated.
    3. The request does not involve a significant reduction in the 
margin of safety.
    The SLMCPR is a Technical Specification numerical value to 
ensure that 99.9% of all fuel rods in the core will avoid transition 
boiling if the limit is not violated. The proposed SLMCPR change 
results from SLMCPR analysis using the accepted methods as 
identified in the Attachment.
    The margin of safety resides between the SLMCPR and the point at 
which fuel fails. Maintaining the MCPR above the proposed SLMCPR 
will maintain the margin of safety associated with GE's SLMCPR 
methodology. Existing plant procedures will continue to ensure that 
the SLMCPR is not violated.
    Therefore, this request does not involve 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 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005.
    NRC Project Director: John N. Hannon.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: December 23, 1998.
    Description of amendment request: The proposed changes will modify 
the Limiting Condition for Operation for Technical Specifications 
3.3.3.7.1 for the chlorine detection system at Waterford Steam Electric 
Station, Unit 3. A change in the alarm/trip setpoint from 3 parts per 
million (ppm) to 2 ppm is requested. Additionally, the proposed request 
corrects a typographical error in Table 3.3-4.
    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. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: The chlorine detection system has no effect on the 
accidents analyzed in Chapter 15 of the Final Safety Analysis 
Report. Its only effect is on habitability of the control room, 
which will be enhanced by specifying a more conservative setpoint in 
the Technical Specifications (TS). Analysis using more conservative 
assumptions show that a setpoint of 2 parts per million (ppm) 
chlorine is acceptable.
    Correcting the typographical error on TS page 3/4 3-19 has no 
effect on the probability or consequences of an accident previously 
evaluated.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: The proposed Technical Specification change in itself 
does not change the design or configuration of the plant. Using a 
more conservative setpoint performs the same function as the old 
setpoint, but it accomplishes this function with increased 
conservatism.
    Correcting the typographical error on TS page 3/4 3-19 will not 
create the possibility of a new or different type of accident from 
any accident previously evaluated.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change

[[Page 9191]]

involve a significant reduction in a margin of safety?
    Response: The chlorine detection system has no effect on a 
margin of safety as defined by Section 2 of the Technical 
Specifications. Its only effect is on habitability of the control 
room, which will be enhanced by a more conservative setpoint 
provided by this change to the Technical Specifications.
    Correcting the typographical error on TS page 3/4 3-19 does not 
involve a significant reduction in a margin of safety.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1400 
L Street N.W., Washington, D.C. 20005-3502.
    NRC Project Director: John N. Hannon.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: January 25, 1999.
    Description of amendment request: The proposed change request will 
modify Technical Specification (TS) 3.5.1 to allow up to 72 hours to 
restore safety injection tank (SIT) to operable status if one SIT is 
inoperable due to boron concentration not within the limits or the 
inability to verify level and pressure. The proposed change would also 
allow up to 24 hours to restore SIT to operable status if one SIT is 
inoperable due to other reasons when Reactor Coolant System pressure is 
greater than or equal to 1750 psia. The ACTIONS for an inoperable SIT 
are being subdivided based on pressurizer pressure to be consistent 
with the current Waterford 3 requirements and applicability. 
Additionally, the Surveillance requirement to sample the SIT after a 1% 
volume increase is being changed to not be required if the source of 
the makeup is the refueling water storage pool. This amendment request 
is a collaborative effort of participating Combustion Engineering 
Owners Group members based on a review of plant operations, 
deterministic and design basis considerations, and plant risk, as well 
as previous generic studies and conclusions drawn by the NRC Staff and 
contained within NUREG-1366, ``Improvements to Technical Specifications 
Surveillance Requirements,'' and NUREG-1432, Revision 1, ``Standard 
Technical Specifications for Combustion Engineering (CE) Plants.'' TS 
Bases 3/4.5.1 will be revised to support above changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: The Safety Injection Tanks (SITs) are passive 
components in the Emergency Core Cooling System. The SITs are not an 
accident initiator in any accident previously evaluated. Therefore, 
this change does not involve an increase in the probability of an 
accident previously evaluated.
    The SITs were designed to mitigate the consequences of Loss of 
Coolant Accidents (LOCA). These proposed changes do not affect any 
of the assumptions used in deterministic LOCA analyses. Hence the 
consequences of accidents previously evaluated do not change.
    In order to fully evaluate the affect of the SIT Allowed Outage 
Time (AOT) extension from 1 hour to 24 hours when one SIT is 
inoperable for reasons other than boron concentration or inability 
to measure level or pressure, probabilistic safety analysis (PSA) 
methods were utilized. The results of these analyses show no 
significant increase in the core damage frequency. As a result, 
there would be no significant increase in the consequences of an 
accident previously evaluated. These analyses are detailed in CE 
NPSD-994, Combustion Engineering Owners Group ``Joint Applications 
Report for Safety Injection Tank AOT/STI Extension.''
    The proposed change to extend the AOT from 1 hour to 72 hours 
when unable to measure level or pressure is acceptable because SIT 
operability is not based on instrumentation availability. Therefore, 
this does not involve a significant increase in the consequences of 
an accident as evaluated and are endorsed by the Nuclear Regulatory 
Commission (NRC) in NUREG-1366, ``Improvements to Technical 
Specifications Surveillance Requirements.'' The inability to measure 
level or pressure is acceptable because the SIT instrumentation 
provides no safety actuation.
    The AOT extension from 1 hour to 72 hours, based upon boron 
concentration outside the prescribed limits does not involve a 
significant increase in the consequences of an accident as evaluated 
and approved by the NRC in NUREG-1432, ``Standard Technical 
Specifications for Combustion Engineering Plants.'' These changes 
are acceptable because the reduced concentration effects on core 
subcriticality during reflood are minor.
    The change in sampling requirements to not require sampling if 
the makeup source is of the same concentration limit as the SIT is 
acceptable as the concentration will remain within the TS limits.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: The proposed change does not alter the design or 
configuration of the plant. It also does not alter the mitigation 
capabilities of any safety system or components. This change 
increases the AOTs for the condition of SIT inoperability. The boron 
concentration is maintained by make-up from a source of water with 
the required concentration of the SITs.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: The proposed changes do not affect the limiting 
conditions for operation or their bases that are used in the 
deterministic analyses to establish the margin of safety. PSA and 
deterministic evaluations were used to evaluate these changes. The 
PSA evaluations demonstrated that the applicable changes are either 
risk neutral or risk beneficial. These evaluations are detailed in 
CE NPSD-994. The deterministic evaluations show that the SITs would 
be able to perform their safety function. These changes are 
consistent with NUREG-1366 and NUREG-1432. The margin of safety is 
not significantly affected by makeup from a source of the same 
concentration limit as the SIT or increase in the AOT for boron 
concentration of one SIT not within limits.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1400 
L Street N.W., Washington, D.C. 20005-3502.
    NRC Project Director: John N. Hannon.

[[Page 9192]]

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: January 25, 1999.
    Description of amendment request: The proposed changes modify 
Technical Specifications Section 6.0 to remove certain administrative 
controls and instead rely on the change controls of 10 CFR 50.54(a)(3) 
and to add a requirement to Section 6.0 concerning the responsibilities 
of the General Manager Plant Operations. The requested changes are 
consistent with the Improved Standard Technical Specifications for 
Combustion Engineering plants, NUREG-1432.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: The requested changes are purely administrative in 
nature. The proposed changes do not affect the operation of any 
structures, systems, or components or the assumptions of any 
accident analyses. The requested changes only affect Section 6.0 of 
the Waterford 3 Technical Specifications which describe the 
administrative controls to be implemented at the site. The requested 
changes either add an additional administrative requirement or 
remove quality assurance program details from the Technical 
Specifications. The details are being removed from the Technical 
Specifications and instead rely on the change controls of 10 CFR 
50.54(a)(3). This submittal makes no changes to the regulatory 
controls governing changes. The requested changes are purely 
administrative in nature.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: The proposed changes to the Technical Specification 
requirements are purely administrative in nature and do not involve 
a change in plant design or affect the configuration or operation of 
any structure, system, or component.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: The proposed changes do not affect the operation of 
any structures, systems, or components or the assumptions of any 
accident analyses. The requested changes are purely administrative 
in nature.

    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety. The NRC staff has reviewed the 
licensee's analysis and, based on this review, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Local Public Document Room Location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1400 
L Street N.W., Washington, D.C. 20005-3502.
    NRC Project Director: John N. Hannon.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa

    Date of amendment request: January 22, 1999.
    Description of amendment request: The proposed amendment would 
revise Duane Arnold Energy Center (DAEC) Technical Specification (TS) 
Section 4.3, ``Fuel Storage,'' by updating the criticality requirements 
(k-infinity and U-235 enrichment limits) for storage of fuel assemblies 
in the spent fuel racks. This change would allow for storage of nuclear 
fuel assemblies with new designs, including GE-12 with a 10X10 pin 
array.
    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:

    After reviewing this proposed amendment, we have concluded:
    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The probability of occurrence of the accident/abnormal 
conditions evaluated in UFSAR Section 9.1.2.3 is not significantly 
increased by this change because no modification in fuel handling 
equipment, fuel pool cooling equipment, fuel storage racks, or fuel 
handling practices is taking place. Only the k-infinity and 
enrichment limits for the stored fuel are being changed.
    The postulated accident/abnormal conditions evaluated in UFSAR 
Section 9.1.2.3 have been re-evaluated for the proposed changes in 
k-infinity and enrichment limits. The results demonstrate that the 
consequences are negligible. The analyses performed show that the 
requirement to maintain K-eff less than 0.95 (substantially 
subcritical) is satisfied for normal and postulated abnormal 
conditions using methods and assumptions that are consistent with 
the existing UFSAR. Seismic adequacy and structural integrity of the 
pool and racks are not affected by the introduction of GE-12 fuel. 
Local and bulk pool temperatures remain bounded by the current UFSAR 
analysis for fuel exposures with GE-12 fuel expected through two 
cycles of operation (i.e., through Cycle 18 operation). Based upon a 
scoping study comparing the hydraulic diameters of GE-10 and GE-12 
fuel, large margins to pool boiling conditions at the final 
discharge exposures of GE-12 fuel will be maintained. Therefore, the 
consequences of the accident are not significantly increased by this 
change.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    No new types of accidents are being introduced because no 
modification in fuel handling equipment, fuel pool cooling 
equipment, fuel storage racks or fuel handling procedures is being 
made. The design basis function of the spent fuel racks is to 
maintain the fuel configuration substantially subcritical and within 
allowable temperatures under both normal and postulated abnormal 
conditions. This design basis function will be maintained with the 
proposed k-infinity and enrichment limits.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    The margin of safety is not significantly reduced. This margin 
is based on the requirement to limit the K-eff of fuel in the spent 
fuel racks to less than 0.95. The proposed changes in k-infinity and 
enrichment limits have been shown to meet this requirement, using 
methods and assumptions that are consistent with the existing UFSAR. 
Seismic adequacy and structural integrity of the pool and racks are 
not affected by the introduction of GE-12 fuel. Local and bulk pool 
temperatures remain bounded by the current UFSAR analysis for fuel 
exposures with GE-12 fuel expected through two cycles of operation 
(i.e., through Cycle 18 operation). Based upon a scoping study 
comparing the hydraulic diameters of GE-10 and GE-12 fuel, large 
margins to pool boiling conditions at the final discharge exposures 
of GE-12 fuel will be maintained.
    Based upon the above, we have determined that the proposed 
amendment will not involve a significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, IA 52401.

[[Page 9193]]

    Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Project Director: Cynthia A. Carpenter.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa

    Date of amendment request: October 15, 1998, as supplemented on 
December 21, 1998.
    Description of amendment request: The proposed amendment would 
revise the Duane Arnold Energy Center (DAEC) Technical Specifications 
(TS) by adding a new TS 3.7.9, ``Control Building/Standby Gas Treatment 
System (CB/SBGT) Instrument Air System.'' The proposed amendment would 
also revise (TS) 3.6.1.3, ``Primary Containment Isolation Valves 
(PCIVs),'' Condition E, by adding a time limit for plant operation if a 
penetration flow path is isolated by a single purge valve with 
resilient seal.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The amendment is adding new requirements for the CB/SBGT 
Instrument Air System that are commensurate with the safety 
functions it supports and consistent with other support systems in 
the Technical Specifications. These requirements provide appropriate 
actions and time limits for plant operation with one or both CB/SBGT 
Instrument Air subsystems inoperable. The probability of an event 
while in this condition is low, and the consequences are bounded by 
the failure of the supported systems. The CB/SBGT Instrument Air 
System is not assumed to be an initiator of an analyzed event.
    The amendment is also adding a time limit for plant operation if 
a purge valve with resilient seal is used to satisfy TS 3.6.1.3 
Required Action E.1 (isolate the affected penetration flow path). 
While primary containment integrity is provided by the purge valve, 
it is prudent to limit operation in this condition due to the 
potential for increased leakage from a single active failure.
    These additions will provide assurance that affected systems 
will be OPERABLE when required and as assumed in the design basis.
    This change will not physically alter the plant (no new or 
different type of equipment will be installed). This change will not 
alter the operation of process variables, structures, systems, or 
components as described in the safety analysis. This change will not 
alter assumptions relative to the mitigation of an accident or 
transient event. This change will not increase the probability of 
initiating, or the consequences of an analyzed event.
    (2) The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The amendment adds new requirements for the CB/SBGT Instrument 
Air System and adds a time limit for plant operation if a purge 
valve with resilient seal is used to satisfy TS 3.6.1.3 Required 
Action E.1.
    This change will not physically alter the plant (no new or 
different type of equipment will be installed). This change will not 
alter the operation of process variables, structures, systems, or 
components as described in the safety analysis. Thus, a new or 
different kind of accident will not be created.
    (3) The proposed amendment will not involve a significant 
reduction in a margin of safety.
    The amendment is adding new requirements for the CB/SBGT 
Instrument Air System to provide appropriate actions and time limits 
for plant operation with one or both CB/SBGT Instrument Air 
subsystems inoperable.
    The amendment is also adding a time limit for plant operation if 
a purge valve with resilient seal is used to satisfy TS 3.6.1.3 
Required Action E.1 (isolate the affected penetration flow path). 
While primary containment integrity is provided by the purge valve, 
it is prudent to limit operation in this condition due to the 
potential for increased leakage from a single active failure in the 
remaining OPERABLE components.
    This change will not physically alter the plant (no new or 
different type of equipment will be installed). This change will not 
alter the operation of process variables, structures, systems, or 
components as described in the safety analysis. This change will not 
alter assumptions relative to the primary success path for 
mitigation of an accident or transient event.
    These additions will provide assurance that the accident 
mitigation functions will perform as assumed in the safety analysis. 
Thus, the margin of safety will not be reduced.

    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: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, IA 52401.
    Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Project Director: Cynthia A. Carpenter.

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

    Date of amendment request: January 29, 1999.
    Description of amendment request: The amendment would revise the 
technical specifications (TS) to relocate three cycle-specific 
parameter limits; shutdown margin with Tcold>210 deg.F, 
moderator temperature coefficient, and minimum boric acid storage tank 
level versus concentration, to the Core Operating Limits Report (COLR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The safety analysis most impacted by a change to the negative 
Moderator Temperature Coefficient (MTC) limit is the Main Steam Line 
Break (MSLB) event. The Steam Line Break Cooldown curves for an MTC 
are calculated and then input to the cycle-specific MSLB analysis 
(if necessary) during the reload analysis process, using an NRC-
approved methodology. The required/acceptable Shutdown Margin (SDM) 
is dependent upon the core loading pattern used (i.e., cycle-
specific core physics parameters) and is largely dependent on the 
cycle-specific MTC and available scram worth. The SDM is determined 
based on the analysis of the Hot Zero Power (HZP) MSLB event in 
which the return-to-critical and return-to-power conditions are 
evaluated to provide acceptable results. With the ongoing changes in 
MTC as a result of core loadings for FCS and higher U-235 
enrichments, the end-of-cycle MTC is becoming more negative than the 
present Technical Specifications limit. Since the MTC is fuel cycle 
specific and influences the required SDM, it is appropriate to move 
both of these values to the COLR, consistent with Generic Letter 88-
16. Note that no change to the SDM for Tcold 
210 deg.F is being proposed.
    The cycle-specific reload analysis is performed for every 
operating cycle and the results, as incorporated into the COLR 
pursuant to the 10 CFR 50.59 process, are transmitted to the NRC. 
FCS will continue to provide COLR updates to the NRC. The relocation 
of the negative MTC and the ``BAST level versus BAST Concentration'' 
curves into the COLR, consistent with the NRC recommendations of 
Generic Letter 88-16, will not modify the methodology used in 
generating the limits, nor the manner in which they are implemented. 
These limits will continue to be determined by analyzing the same 
postulated events as previously analyzed. FCS will continue to 
operate within the limits specified in the COLR and will take the 
same corrective actions when or if these limits are exceeded as 
required by

[[Page 9194]]

current Technical Specifications. The potential increase of the 
absolute magnitude of the negative MTC with Shutdown Margin decrease 
is evaluated during the COLR reload analysis process in accordance 
with OPPD's NRC-approved topical report. Therefore, this proposed 
amendment is administrative in nature and has been concluded not to 
increase 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 changes to FCS Technical Specifications were the 
result of a recommendation from a Generic Letter. Future changes to 
the parameters being relocated to the COLR can only be performed 
with approved Reload Analyses. No new or different kind of accident 
is created by this administrative change because the actual 
operation of FCS remains unchanged. Therefore the possibility of an 
accident or malfunction of a different type than previously 
evaluated in the safety analysis report would not be created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    As indicated above, the implementation of this proposed COLR 
change, consistent with the guidance of Generic Letter 88-16, makes 
use of the existing safety analysis methodologies and the resulting 
limits and setpoints for plant operation. Additionally, the safety 
analysis acceptance criteria for operation with this proposed 
amendment have not changed from the criteria used in the current 
reload analysis. Therefore, the margin of safety as defined in the 
bases of Technical Specifications is not reduced.

    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: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: January 12, 1999.
    Description of amendment request: The proposed change involves 
revising Technical Specification (TS) Section 3/4.4.2, ``Safety/Relief 
Valves,'' and TS Bases Sections B 3/4.4.2, B 3/4.5.1 and B 3/4.5.2, to 
increase the allowable as-found main steam Safety Relief Valve (SRV) 
code safety function lift setpoint tolerance from plus or minus 1% to 
plus or minus 3%. This change will also require the as-left SRV code 
safety function lift setting to be set within plus or minus 1% of the 
specified nominal lift setpoint prior to reinstallation in the plant. 
In support of this proposed TS change, the required number of OPERABLE 
SRVs in Operational Conditions (OPCONs) 1, 2, and 3 will be changed 
from 11 to 12. The number of SRVs in each lift pressure grouping will 
remain the same. This proposed TS change does not alter the SRV nominal 
lift setpoints or the SRV lift setpoint test frequency currently 
specified by TS Section 3/4.4.2. The proposed change does not change 
the SRV testing commitment specified in LGS Updated Final Safety 
Analysis Report (UFSAR) Chapter 5.2.2.10, ``Inspection and Testing.''
    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 Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed TS changes allow for an increase in the as-found 
main steam Safety Relief Valve (SRV) setpoint tolerance from plus or 
minus 1% to plus or minus 3%. The proposed changes also reduce the 
allowable number of SRVs to be out-of-service from three (3) to two 
(2). The proposed changes do not alter the SRV nominal lift 
setpoints or SRV lift setpoint test frequency. The actuation of an 
SRV is the precursor to the inadvertent opening of a SRV transient, 
as discussed in Updated Final Safety Analysis Report (UFSAR) Chapter 
15.1.4. Increasing the allowable as-found SRV code safety function 
lift setpoint tolerance from plus or minus 1% to plus or minus 3% 
does have the potential for the minimum SRV simmer margin to be 
reduced from 113.3 psig to 89.9 psig. A reduction in simmer margin 
will not directly result in an increase of the probability on an 
inadvertent self actuation of an SRV. A reduction in simmer margin 
will reduce the seating force which may initiate leakage. However, 
this leakage is monitored and corrective actions can be implemented 
prior to progressing to the point of the potential of an inadvertent 
actuation. This reduction in SRV simmer margin has been evaluated by 
the SRV manufacturer and determined to be acceptable; therefore, the 
probability of an inadvertent SRV actuation remains unchanged. 
Actuation of an SRV is not a precursor for any other event evaluated 
in the Safety Analysis Report (SAR).
    The proposed TS changes have been evaluated on both a generic 
and plant specific basis. The NRC has approved the general approach 
of this change; however, implementation is contingent on several 
plant specific evaluations. The required plant specific analyses and 
evaluations included transient analysis of the anticipated 
operational transients (AOTs); analysis of the design basis 
overpressurization event; evaluation of the performance of high 
pressure systems, motor operated valves, and vessel instrumentation 
and associated piping; and evaluation of the containment response 
during Loss-of-Coolant Accident (LOCA) and hydrodynamic loads on the 
SRV discharge lines and containment. In addition to the plant 
specific analyses and evaluations required by the NRC, the following 
items were also considered: ECCS/LOCA [Emergency Core Cooling 
System] performance, SRV simmer margin, high pressure--low pressure 
interfaces, i.e., High Energy Line Break (HELB), Station Blackout 
(SBO), and Fire Safe Shutdown (FSSD), and the short term 
pressurization phase of an ATWS [anticipated transient without 
scram] event. These analyses and evaluations show that there is 
adequate margin to the design core thermal limits and reactor vessel 
pressure limits using the plus or minus 3% SRV code safety function 
lift setpoint tolerance and two (2) SRVs out-of-service. The 
analyses and evaluations also show that the operation of the high 
pressure injection systems will not be adversely affected, that SRV 
discharge piping stresses will not be exceeded, and that the 
containment response during a LOCA will be acceptable.
    Evaluations of the impact of the proposed change on the 
Equipment Important to Safety have been performed and no adverse 
conditions were identified. The reactor pressure vessel and attached 
systems and piping have been evaluated for the impact of this 
proposed TS change. A plant specific analysis has been performed 
which indicates that neither the American Society of Mechanical 
Engineers (ASME) Code upset limits or the TS Safety Limits for the 
reactor pressure vessel will be exceeded for the limiting event, 
i.e., Main Steam Isolation Valve (MSIV) closure with flux Scram. The 
reactor pressure vessel and attached piping design values will not 
be exceeded. The current high pressure--low pressure interface 
evaluation utilized nominal SRV setpoints, and therefore, is 
unaffected. Therefore, the probability of a malfunction of the 
reactor pressure vessel and attached systems and piping is not 
increased.
    The nuclear fuel has been evaluated for the impact of the 
proposed change. Plant specific analyses were performed which 
indicate that for all abnormal operational transients adequate 
margin to the limiting thermal limit parameter, i.e., Minimum 
Critical Power Ratio (MCPR), is maintained. Emergency Core Cooling 
System (ECCS)/LOCA performance is maintained adequate to meet the 
requirements of 10CFR50.46. Therefore, the probability of the 
malfunction of the nuclear fuel is not increased.
    The SRVs have been evaluated for the impact of the proposed TS 
changes. No physical changes to the SRVs will be made as a result of 
the proposed TS changes. Adequate simmer margin will be maintained 
with the increased tolerance to ensure that an inadvertent lifting 
of a SRV does not occur.

[[Page 9195]]

The increase in SRV discharge flow and reactor vessel pressure due 
to the potential for higher SRV lift setpoints are bounded by the 
SRV steam flows and reactor vessel pressure currently used in the 
evaluation of SRV discharge piping, quencher, quencher support, and 
hydrodynamic loads on the suppression pool and submerged structures; 
therefore, the probability of a malfunction of a SRV or associated 
components and structures is not increased.
    The Containment response during a LOCA has been evaluated for 
the impact of the proposed change. The major factor in the 
Containment response to a LOCA is the rate of reactor vessel water 
inventory loss. The rate of reactor vessel water inventory loss is 
mainly dependent on reactor decay heat which is not affected by the 
proposed change. Therefore, the probability of the malfunction of 
the Containment is not increased.
    The High Pressure Coolant Injection (HPCI) system has been 
evaluated for the impact of the proposed TS changes. The analysis 
determined that the HPCI system would not be capable of developing 
its design flowrate of 5600 gpm at a reactor pressure of 1205 psig 
(lowest SRV nominal setpoint +3% tolerance) unless the HPCI turbine/
pump maximum rated speed was increased. However, increasing the HPCI 
turbine/pump maximum rated speed is prevented due to HPCI pump 
discharge piping overpressurization concerns. Further analysis has 
shown that the HPCI system is capable of meeting its required ECCS 
function design flowrate, and its required non-ECCS flowrate, 
without any change to the current system operating parameters. 
Therefore, the probability of a malfunction of the HPCI System is 
not increased.
    The Reactor Core Isolation Cooling (RCIC) system has been 
evaluated for the impact of the proposed change. The analysis 
determined that in order for the RCIC system to be capable of 
injecting its design flowrate of 600 gpm at a reactor pressure of 
1205 psig (lowest SRV setpoint of 1170 psig +3% tolerance) the 
maximum rated speed of the RCIC turbine/pump is required to be 
increased from 4575 rpm to 4625 rpm. This increase in the RCIC 
turbine/pump maximum rated speed will reduce the margin to the 
overspeed trip from 123% to 122.1%. This reduction in the margin to 
the overspeed trip is acceptable due to the implementation of plant 
Modification P00210, ``RCIC System Startup Transient Improvement,'' 
which reduced the amount of turbine/pump speed overshoot during 
system startup. The RCIC overspeed trip setpoint will not be 
changed; therefore, a failure of the RCIC turbine/pump (missile 
hazard or system overpressurization) due to overspeed is not 
increased. All other RCIC System components will continue to operate 
within the currently specified design and operating limits. 
Therefore, the probability of a malfunction of the RCIC System is 
not increased.
    The Standby Liquid Control (SLC) system has been evaluated for 
the impact of the proposed change. The SLC system capability of 
shutting down the reactor during a postulated event in which all or 
some of the control rods cannot be inserted or during a postulated 
Anticipated Transient Without Scram (ATWS) event is not impacted by 
this proposed change. Therefore, the probability of a malfunction of 
the SLCS is not increased.
    The Control Rod Drive (CRD) system has been evaluated for the 
impact of the proposed change. The CRD system capability of 
controlling reactor power during normal plant operation and rapidly 
inserting control rod blades (Scram) during abnormal plant 
conditions is not impacted by the proposed change. Therefore, the 
probability of a malfunction of the CRD system is not increased.
    The Reactor Vessel Instrumentation System has been evaluated for 
the impact of the proposed change. The Reactor Vessel 
Instrumentation System will continue to be operated within the 
current design pressure/temperature requirements; therefore, the 
probability of a malfunction of the Reactor Vessel Instrumentation 
System is not increased.
    The LGS, Units 1 and 2, Generic Letter 89-10 Motor-Operated 
Valve (MOV) Program has been evaluated for the proposed change. The 
LGS MOV Program currently uses SRV nominal setpoints for 
differential pressure determinations for valves in which reactor 
pressure at the SRV setpoint is limiting. Use of nominal SRV 
setpoints is consistent with current industry practice. Therefore, 
the probability of a malfunction of a MOV is not increased.
    Reducing the number of SRVs allowed to be out-of-service does 
not make the consequences of a malfunction of a SRV more severe, 
since the number of SRVs required to maintain the reactor vessel 
within ASME Code and TS Safety Limits will be maintained OPERABLE. 
The proposed change does not result in any changes to the 
interactions of any system, structure, or component. All systems, 
structures, and components will continue to function as designed.
    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 allow for an increase in the as-found 
SRV setpoint tolerance from plus or minus 1% to plus or minus 3%. 
The proposed TS changes also reduce the allowable number of SRVs to 
be out-of-service from three (3) to two (2). Generic and plant 
specific analyses and evaluations indicate that the plant response 
to any previously evaluated event will remain unchanged. All plant 
systems, structures, and components will continue to be capable of 
performing their required safety function as required by event 
analysis guidance.
    The proposed TS changes do not alter the SRV nominal lift 
setpoints or SRV lift setpoint test frequency. The operation and 
response of the affected Equipment Important to Safety is unchanged. 
All systems, structures, and components will continue to be operated 
within acceptable operating and/or design parameters. No system, 
structure, or component will be subjected to a condition that has 
not been evaluated and determined to be acceptable using the 
guidance required for specific event analysis.
    Therefore, the proposed TS changes do not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed TS changes allow for an increase in the as-found 
SRV setpoint tolerance from plus or minus 1% to plus or minus 3%. 
The proposed TS changes also reduce the allowable number of SRVs to 
be out-of-service from three (3) to two (2). The proposed TS changes 
do not alter the SRV nominal lift setpoints or SRV lift setpoint 
test frequency. The operation and response of the affected Equipment 
Important to Safety is unchanged. All systems, structures, and 
components will continue to be operated within acceptable operating 
and/or design parameters. While the calculated peak reactor vessel 
pressure for the ASME overpressure event and the ATWS Pressure 
Regulator Failure-Open (PREGO) event are higher than those 
calculated without the increase in setpoint tolerance, both are 
still within the respective licensing acceptance limits associated 
with these events. These licensing acceptance limits have been 
determined by the NRC to provide a sufficient margin of safety.
    The increase in the RCIC system turbine/pump maximum rated speed 
is within the capability of the system design. The reduction in the 
margin to the overspeed trip is not a reduction in the margin of 
safety, since the operation of the RCIC System has demonstrated 
minimal speed overshoot on system initiation due to the installation 
of plant Modification P00210, ``RCIC System Startup Transient 
Improvement.''
    The inability of the HPCI system to be capable of injecting 5600 
gpm at a reactor pressure of 1205 psig (lowest SRV nominal setpoint 
of 1170 psig +3% tolerance) is not a reduction in the margin of 
safety, since analysis for events that would result in high reactor 
vessel pressure indicate that the HPCI System is capable of 
providing adequate coolant injection.
    The increase in SRV steam flow and reactor vessel pressure does 
not reduce the margin of safety associated with the SRVs and 
associated components and structures since the increased SRV steam 
flow rate and reactor vessel pressure are bounded by the current 
design analysis.
    The margin of safety for fuel thermal limits and 10CFR50.46 
limits is unaffected by the proposed change.
    The margin of safety for the Containment is unaffected by the 
proposed change.
    The capability of the SLC system to perform its safety function 
during all required events, using the required guidance for event 
analysis, is maintained. Therefore, the proposed changes do not 
reduce the margin of safety provided by the SLC system.
    Therefore, these proposed TS changes do not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this

[[Page 9196]]

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: 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 Project Director: William M. Dean.

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

    Date of amendment request: January 25, 1999.
    Description of amendment request: The proposed Technical 
Specification (TS) Change Request revises the TS Surveillance 
Requirement frequencies for Sections 4.8.1.1.2.e.1, 4.8.1.1.2.e.8.a, 
and 4.8.1.1.2.e.8.b for the Emergency Diesel Generator maintenance 
inspection outages, the 24-hour endurance run, and for the hot restart 
test from 18 to 24 months.
    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 Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The maintenance inspection interval change and the corresponding 
interval change for the associated 24 hour endurance test and hot 
restart test which are normally performed in conjunction with the 
diesel preventive maintenance overhaul inspections, as well as the 
programmatic improvements addressed here do not involve physical 
changes that would affect the ability of the EDGs [emergency diesel 
generators] to perform their safety function. The Emergency Diesel 
Generator System is not an accident initiator.
    The Surveillance Testing requirements of Technical Specification 
Section 3/4.8 will continue to verify the operability and 
reliability of the Emergency Diesel Generator system.
    The proposed changes do not affect the ability of the EDGs to 
mitigate the consequences of an accident, including the Loss of 
Coolant Accident (LOCA) coupled with Loss Of Offsite Power accident 
analyses as presented in Chapter 15 of the LGS [Limerick Generating 
Station] UFSAR [Updated Final Safety Analysis Report]. EDG 
unavailability due mostly to outage inspections is more than 2 times 
higher than EDG unplanned unavailability. An extension of the outage 
inspection frequency to 24 months will result in increased EDG 
availability to mitigate the consequences of a potential accident. 
When this program is taken in its entirety the extended maintenance 
intervals coupled with the defined enhancements is judged to result 
in an overall increase in EDG availability and reliability. 
Therefore, the probability or consequences of an accident previously 
evaluated is not increased.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The Emergency Diesel Generator system is not an accident 
initiator. The operation and design of the onsite emergency power 
system (including the EDGs) is not being changed; only the overhaul 
inspection interval coupled with the program improvements and the 
corresponding interval change for the associated 24 hour endurance 
test and hot restart test, (which are normally performed in 
conjunction with the diesel preventive maintenance overhaul 
inspections), are changed. The EDG system meets the single failure 
criteria at the EDG unit level, i.e., the SAR [safety analysis 
report] states that with one EDG failed or out-of-service, the 
standby AC system is capable of furnishing sufficient power for the 
minimum Class 1E load demand, assuming a limiting design basis 
accident has occurred. The proposed changes involve a routine 
preventive maintenance and inspection time interval change along 
with the corresponding surveillance test interval changes, and also 
include programmatic improvements to reduce the likelihood of a 
failure of an individual EDG unit; the proposed changes do not 
involve any physical design or operational changes that could create 
a malfunction extending beyond an individual EDG nor do they 
increase the potential for a common-mode EDG failure. Therefore, it 
is not possible to create a new or different type of accident 
through implementation of these changes.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The changes to bring the frequencies of the EDG overhaul, the 24 
hour endurance test and the associated hot restart test into 
alignment with the current 2 year operating cycle, and the detailed 
programmatic changes to achieve conformance with the FMOG [Fairbanks 
Morse Owners Group] recommended maintenance program, will increase 
the reliability and availability of the EDG system. This will 
enhance the margin of safety as the amount of time the EDGs are out-
of-service will decrease and the system will be single-failure proof 
for more clock hours when the nuclear reactor(s) are operating. The 
changes discussed here do not result in operation of the emergency 
diesel generator system nor any other plant system in a manner 
beyond their original design basis, and thus does not reduce any 
explicit or implicit Technical Specification 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: 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 Project Director: William M. Dean.

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

    Date of application for amendment: February 12, 1997.
    Brief description of amendment: The proposed amendment would delete 
a portion of the Trojan site from the 10 CFR 50 license when that 
portion of the site, designated for use as an independently licensed 
spent fuel storage installation (ISFSI), receives a part 72 license.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensees' analysis 
against the standards of 10 CFR 50.92(c). The licensee's analysis is 
summarized below:
    The proposed changes would not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed change is administrative in nature and has no impact on 
the probability or consequences of accidents previously evaluated. The 
physical structures, systems, and components of the Trojan Nuclear 
Plant and the operating procedures for their use are unaffected by this 
proposed change. The proposed action would eliminate the ISFSI area 
from the Part 50 license when the Part 72 license is issued. The 10 CFR 
72 licensing controls for the area will assure an adequate level of 
safety for the area during normal operation of the ISFSI and during 
abnormal events or accidents. Therefore the proposed Part 50 amendment 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes would not create the possibility of a new or 
different kind of accident from any accident previously evaluated. The 
proposed action would eliminate the ISFSI area from the Part 50 license 
when the Part 72 license is issued. The proposed change is 
administrative in

[[Page 9197]]

nature and has no impact on plant systems, structures, or components or 
on any procedures for operating the plant equipment. The ISFSI will be 
separately licensed under Part 72 and physically separated from the 
Part 50 licensed structures and equipment. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from those previously evaluated.
    The proposed changes do not involve reduction in the margin of 
safety. The Trojan Permanently Defueled Technical Specifications (PDTS) 
contain four limiting conditions of operation that address: 1) Spent 
Fuel Water Level, 2) Spent Fuel Pool Boron Concentration, 3) Spent Fuel 
Pool Temperature, and 4) Spent Fuel Pool load restrictions. These PDTS 
will remain in effect as long as spent fuel is stored in the Spent Fuel 
Pool, which is in accordance with their applicability statements. The 
ISFSI area is physically separated from the Spent Fuel Pool area and 
the Fuel Building and will have no effect on spent fuel water level, 
spent fuel pool boron concentration, spent fuel pool temperature, or 
loads over the Spent Fuel Pool. The proposed change is administrative 
and does not affect plant equipment, operating parameters, or 
procedures. Based on the above, the proposed change will not reduce the 
margin of safety.
    Based on a staff review of the licensee's analysis, 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: Branford Price Millar Library, 
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151, 
Portland, Oregon 97207.
    Attorney for licensees: Leonard A. Girard, Esq., Portland General 
Electric Company, 121 S. W. Salmon Street, Portland, Oregon 97204.
    NRR Project Director: Seymour H. Weiss.

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

    Date of application for amendment: January 7, 1999.
    Brief description of amendment: The proposed amendment would allow 
loading and handling of spent fuel transfer and storage casks in the 
Trojan Fuel Building.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensees' analysis 
against the standards of 10 CFR 50.92(c). The licensee's analysis is 
summarized below:
    The proposed changes would not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
With the permanent cessation of operations, the number of potential 
accidents was reduced to those types of accidents associated with the 
storage of irradiated fuel and radioactive waste storage and handling. 
Additional events were postulated for decommissioning activities due to 
the difference in the types of activities that were to be performed. 
The postulated accidents in the Defueled Safety Analysis Report (DSAR) 
are generally classified as: (1) radioactive release from a subsystem 
or component, (2) fuel handling accident and, (3) loss of spent fuel 
decay heat removal capability. The postulated events described in the 
Decommissioning Plan are grouped as: (1) decontamination, 
dismantlement, and materials handling events, (2) loss of support 
systems (offsite power, cooling water, and compressed air), (3) fire 
and explosions, and (4) external events (earthquake, external flooding, 
tornadoes, extreme winds, volcanoes, lightning, toxic chemical 
release). These types of accidents are discussed below.
    Radioactive release from a subsystem or component involves failure 
of a radioactive waste gas decay tank (WGDT) or failure of a chemical 
and volume control system holdup tank (HUT). For a failure of a WGDT, 
the radioactive contents are assumed to be principally the noble gases 
krypton and xenon, the particulate daughters of some of the krypton and 
xenon isotopes and trace quantities of halogens. For the failure of a 
HUT, the assumptions were full power operations with 1-percent failed 
fuel, 40 weeks elapsed since power operation, and 60,000 gallons of 
120 deg. F liquid released over a 2-hour period. However, the WGDT's 
and HUT's are no longer active and have been emptied. Therefore, cask 
loading and transfer activities cannot increase the probability of 
occurrence of a failure or the consequence of a failure of the WGDT's 
or HUT's.
    The fuel handling accident involves a stuck or dropped fuel 
assembly that results in damage of the cladding of the fuel rods in one 
assembly and the release of gaseous fission products. Spent fuel 
handling and loading will involve moving the spent fuel assemblies one 
by one, from the Spent Fuel Pool to the baskets which will be located 
in the Cask Loading Pit. The fuel handling equipment will be the same 
as had been previously analyzed with the exception of special tools 
which will be used to manipulate failed fuel. These special tools will 
be similar in size and weight to the existing tools used for underwater 
manipulation and therefore will not present a new hazard. In addition, 
the same administrative controls and physical limitations imposed on 
any fuel handling operation will be used for spent fuel loading and 
handling. The potential release, 100 percent of gap noble gas, from a 
fuel assembly is not affected (although the fission product inventory 
in a fuel assembly continues to decrease with time). Thus there is no 
increase in the probability of occurrence or consequences of a fuel 
handling accident over what would be expected for any routine fuel 
handling operation.
    The loss of spent fuel decay heat removal capability involves the 
loss of forced spent fuel cooling with and without concurrent Spent 
Fuel Pool inventory loss. The only requirement to assure adequate decay 
heat removal capability for the spent fuel is to maintain the water 
level in the Spent Fuel Pool so that the fuel assemblies remain covered 
(i.e. the capability to make up water to the Spent Fuel Pool must be 
available when required). The potential events which could result in a 
loss of spent fuel decay heat removal include external events 
(explosions, toxic chemical, fires, ship collision with intake 
structure, oil or corrosive liquid spills in the river, cooling tower 
collapse, seismic events, severe meteorological events), and internal 
events including Spent Fuel Pool makeup water system malfunctions 
(Service Water System, electrical power, instrument air). Spent fuel 
loading and handling will not require the use of explosive materials 
(the gases used for electric arc welding are inert), toxic chemicals or 
flammable materials (routine use of contamination control materials is 
not considered to present a significant hazard). The probability of 
other external events (e.g. cooling tower collapse) is not effected by 
the spent fuel handling and loading activities inside the Fuel 
Building. Spent fuel loading and handling activities will not directly 
interface with the Spent Fuel Pool makeup water systems, therefore does 
not affect their probability of failure. (The Cask Loading Pit will be 
filled with borated water from the Spent Fuel Pool that will be cooled 
by the Spent Fuel Cooling System, but use of this water in the Cask 
Loading Pit does not increase the failure probability of

[[Page 9198]]

the Spent Fuel Pool or makeup water systems.) As described in the 
licensees' safety evaluation, the safe load path and handling height 
limitations will ensure that a load drop does not adversely affect the 
Spent Fuel Pool or the makeup water systems. Therefore there is no 
significant increase in the probability or consequences of a loss of 
spent fuel decay heat removal capability.
    The events postulated in the Decommissioning Plan are similar to 
the DSAR with the exception of the decontamination, dismantlement, and 
materials handling events. Decontamination events involve gross liquid 
leakage from in-situ decontamination equipment (e.g. tanks) or 
accidental spraying of liquids containing concentrated contamination. 
Dismantlement events involve segmentation of components and structures, 
or removal of concrete by rock splitting, explosives, or electric and/
or pneumatic hammers. Dismantlement events potentially result in 
airborne contamination. Material handling events involve the dropping 
of contaminated components, concrete rubble, filters, or packages of 
particulate materials. Licensee administrative controls will be 
implemented to ensure that spent fuel loading and handling activities 
and decommissioning activities will not be performed concurrently if 
they interact with each other and could increase the probability or 
consequences of a postulated event of accident. Therefore, neither the 
probability nor the consequences of decontamination, dismantlement, and 
materials handling events will not be significantly increased.
    The proposed changes would not create the possibility of a new or 
different kind of accident from any accident previously evaluated. As 
described in the licensees' safety evaluation the potential accidents 
associated with fuel handling and loading were similar to fuel handling 
accidents, material handling events and pressurized line break 
previously analyzed. Additionally the potential consequences were a 
small fraction of Environmental Protection Agency (EPA) Protective 
Action Guides (PAG's). Therefore, fuel loading and handling does not 
present new or different types of accidents.
    The proposed changes do not involve a significant reduction in the 
margin of safety. The Trojan Permanently Defueled Technical 
Specifications (PDTS) contain four limiting conditions of operation 
that address: (1) Spent fuel water level, (2) spent fuel pool boron 
concentration, (3) spent fuel pool temperature, and (4) spent fuel pool 
load restrictions. These PDTS will remain in effect as long as spent 
fuel is stored in the Spent Fuel Pool, which is in accordance with 
their applicability statements. The spent fuel loading and handling 
activities will not affect these PDTS or their bases.
    The Cask Loading Pit, where the spent fuel will be loaded into the 
basket, is immediately adjacent to the Spent Fuel Pool. The gate 
between the Cask Loading Pit and Spent Fuel Pool will be open to allow 
transfer of spent fuel assemblies from storage racks in the Spent Fuel 
Pool to the basket in the Cask Loading Pit. Opening the gate between 
them will allow free exchange of water between the Cask Loading Pit and 
the Spent Fuel Pool. The Cask Loading Pit will be filled with borated 
water at approximately the same concentration and temperature as the 
Spent Fuel Pool prior to opening the gate. This will maintain the 
limiting conditions for operation for Spent Fuel Pool boron 
concentration, temperature, and water level and the margin of safety 
will not be affected.
    Spent fuel loading and handling activities will involve lifting and 
moving heavy loads (e.g. transfer cask, basket). Loads that will be 
carried over fuel in the Spent Fuel Pool racks and the heights at which 
they will be carried will be limited to preclude impact energies over 
240,000 in-lbs if the loads were dropped. This is in accordance with 
limiting condition for operation 3.1.4 ``Spent Fuel Pool Load 
Restrictions.'' With this precaution, the limiting condition for 
operation pertaining to load restrictions over the Spent Fuel Pool will 
be satisfied and the margin of safety will be unaffected. The safe load 
paths for heavy loads being lifted outside the Spent Fuel Pool will be 
sufficiently far from the Spent Fuel Pool so as to not have an 
interaction in the unlikely event of a load drop. In addition 
mechanical stops and electrical interlocks on the Fuel Building 
overhead crane will provide additional assurance that heavy loads are 
not carried over the Spent Fuel Pool racks.
    Based on the above, the spent fuel loading and handling activities 
will not reduce the margin of safety.
    Based on a staff review of the licensee's analysis, 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: Branford Price Millar Library, 
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151, 
Portland, Oregon 97207.
    Attorney for licensees: Leonard A. Girard, Esq., Portland General 
Electric Company, 121 S. W. Salmon Street, Portland, Oregon 97204.
    NRR Project Director: Seymour H. Weiss.

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

    Date of application for amendment: January 27, 1999.
    Brief description of amendment: The proposed amendment would allow 
unloading of spent fuel transfer casks in the Trojan Fuel Building.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their 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 licensee's analysis is 
summarized below:
    The proposed changes would not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
With the permanent cessation of operations, the number of potential 
accidents was reduced to those types of accidents associated with the 
storage of irradiated fuel and radioactive waste storage and handling. 
Additional events were postulated for decommissioning activities due to 
the difference in the types of activities that were to be performed. 
The postulated accidents in the Defueled Safety Analysis Report (DSAR) 
are generally classified as: (1) Radioactive release from a subsystem 
or component, (2) fuel handling accident and, (3) loss of spent fuel 
decay heat removal capability. The postulated events described in the 
Decommissioning Plan are grouped as: (1) Decontamination, 
dismantlement, and materials handling events, (2) loss of support 
systems (offsite power, cooling water, and compressed air), (3) fire 
and explosions, and (4) external events (earthquake, external flooding, 
tornadoes, extreme winds, volcanoes, lightning, and toxic chemical 
release). These types of accidents are discussed below.
    Radioactive release from a subsystem or component involves failure 
of a radioactive waste gas decay tank (WGDT) or failure of a chemical 
and volume control system holdup tank (HUT). For a failure of a WGDT, 
the radioactive contents are assumed to be principally the noble gases 
krypton and xenon, the particulate daughters of some

[[Page 9199]]

of the krypton and xenon isotopes and trace quantities of halogens. For 
the failure of a HUT, the assumptions were full power operations with 
1-percent failed fuel, 40 weeks elapsed since power operation, and 
60,000 gallons of 120 deg. F liquid released over a two hour period. 
However, the WGDT's and HUT's are no longer active and have been 
emptied. Therefore, cask loading and transfer activities cannot 
increase the probability of occurrence of a failure or the consequence 
of a failure of the WGDT's or HUT's.
    The fuel handling accident involves a stuck or dropped fuel 
assembly that results in damage of the cladding of the fuel rods in one 
assembly and the release of gaseous fission products. Spent fuel cask 
unloading will involve moving the spent fuel assemblies one by one, 
from the baskets which will be located in the cask loading pit to the 
spent fuel pool. The fuel handling equipment will be the same as had 
been previously analyzed. In addition, the same administrative controls 
on physical limitations imposed on fuel handling and fuel loading 
operations will be used for fuel unloading. The potential release, 100 
percent of noble gases within the gap, from a fuel assembly is not 
affected (although the inventory in a radioactive stored fuel assembly 
continues to decrease with time). Thus, there is no increase in the 
probability of occurrence or consequences of a fuel handling accident 
over what would be expected for any routine fuel handling operation or 
loading of fuel into a cask.
    The loss of spent fuel decay heat removal capability involves the 
loss of forced spent fuel cooling with and without concurrent spent 
fuel pool inventory loss. The only requirement to assure adequate decay 
heat removal capability for the spent fuel is to maintain the water 
level in the spent fuel pool so that the fuel assemblies remain covered 
(i.e., the capability to make up water to the spent fuel pool must be 
available when required). The potential events that could result in a 
loss of spent fuel decay heat removal include external events 
(explosions, toxic chemical, fires, ship collision with intake 
structure, oil or corrosive liquid spills in the river, cooling tower 
collapse, seismic events, and severe meteorological events), and 
internal events including spent fuel pool makeup water system 
malfunctions (service water system, electrical power, and instrument 
air). Spent fuel cask unloading will not require the use of explosive 
materials, toxic chemicals or flammable materials (routine use of 
contamination control materials is not considered to present a 
significant hazard). The probability of other external events (e.g. 
cooling tower collapse) is not effected by the spent fuel unloading 
activities inside the fuel building. Spent fuel cask unloading 
activities will not directly interface with the spent fuel pool makeup 
water systems, and therefore does not affect their probability of 
failure. (The cask loading pit will be filled with borated water from 
the spent fuel pool that will be cooled by the spent fuel cooling 
system, but use of this water in the cask loading pit does not increase 
the failure probability of the spent fuel pool or makeup water 
systems). As described in the licensees' safety evaluation, the safe 
load path and handling height limitations will ensure that a load drop 
does not adversely affect the spent fuel pool or the makeup water 
systems. Therefore, there is no significant increase in the probability 
or consequences of a loss of spent fuel decay heat removal capability.
    The events postulated in the Decommissioning Plan are similar to 
the DSAR with the exception of the decontamination, dismantlement, and 
materials handling events. Decontamination events involve gross liquid 
leakage from in-situ decontamination equipment (e.g. tanks) or 
accidental spraying of liquids containing concentrated contamination. 
Dismantlement events involve segmentation of components and structures, 
or removal of concrete by rock splitting, explosives, or electric and/
or pneumatic hammers. Dismantlement events potentially result in 
airborne contamination. Material handling events involve the dropping 
of contaminated components, concrete rubble, filters, or packages of 
particulate materials. Licensee administrative controls will be 
implemented to ensure that spent fuel cask unloading activities and 
decommissioning activities will not be performed concurrently if they 
interact with each other and could increase the probability or 
consequences of a postulated event of accident. Therefore, neither the 
probability nor the consequences of decontamination, dismantlement, and 
materials handling events will be significantly increased.
    The proposed changes would not create the possibility of a new or 
different kind of accident from any accident previously evaluated. As 
described in the licensee's safety evaluation the potential accidents 
associated with fuel cask unloading were similar to fuel handling 
accidents, material handling events and pressurized line break 
previously analyzed. Additionally the potential consequences were a 
small fraction of Environmental Protection Agency (EPA) Protective 
Action Guides (PAGs). Therefore, fuel loading and handling does not 
present new or different types of accidents.
    The proposed changes do not involve a significant reduction in the 
margin of safety. The Trojan Permanently Defueled Technical 
Specifications (PDTS) contain four limiting conditions of operation 
that address: (1) spent fuel pool water level, (2) spent fuel pool 
boron concentration, (3) spent fuel pool temperature, and (4) spent 
fuel pool load restrictions. These PDTS will remain in effect as long 
as spent fuel is stored in the spent fuel pool, which is in accordance 
with their applicability statements. The spent fuel cask unloading 
activities will not affect these PDTS or their bases.
    The cask loading pit, where the spent fuel will be unloaded from 
basket, is immediately adjacent to the spent fuel pool. The gate 
between the cask loading pit and spent fuel pool will be open to allow 
transfer of spent fuel assemblies from the basket in the cask loading 
pit to the storage racks in the spent fuel pool. Opening the gate 
between them will allow free exchange of water between the cask loading 
pit and the spent fuel pool. The cask loading pit will be filled with 
borated water at approximately the same concentration and temperature 
as the spent fuel pool prior to initial cask loading. This will 
maintain the limiting conditions for operation for spent fuel pool 
boron concentration, temperature, and water level and the margin of 
safety will not be affected.
    Spent fuel cask unloading activities may involve lifting and moving 
heavy loads (e.g. transfer cask, basket). Loads that will be carried 
over fuel in the spent fuel pool racks and the heights at which they 
will be carried will be limited to preclude impact energies over 
240,000 in-lbs if the loads were dropped. This is in accordance with 
limiting condition for operation 3.1.4 ``Spent Fuel Pool Load 
Restrictions.'' With this precaution, the limiting condition for 
operation pertaining to load restrictions over the spent fuel pool will 
be satisfied and the margin of safety will be unaffected. The safe load 
paths for heavy loads being lifted outside the spent fuel pool will be 
sufficiently far from the spent fuel pool so as to not have an 
interaction in the unlikely event of a load drop. In addition, 
mechanical stops and electrical interlocks on the fuel building 
overhead crane will provide additional assurance that heavy loads are 
not carried over the spent fuel pool racks.

[[Page 9200]]

    Based on the above, the spent fuel cask unloading activities will 
not reduce the margin of safety.
    Based on a staff review of the licensee's analysis, 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: Branford Price Millar Library, 
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151, 
Portland, Oregon 97207.
    Attorney for licensees: Leonard A. Girard, Esq., Portland General 
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
    NRR Project Director: Seymour H. Weiss.

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

    Date of amendment request: October 16, 1998, as supplemented 
January 28, 1999.
    Description of amendment request: This application for amendment to 
the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to 
relocate the Chemical Volume and Control System (CVCS) TS 3.2 from the 
TSs to the IP3 Operational Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously analyzed?
    Response: Relocation (i.e., removal from TS) of TS 3.2, the 
bases and the associated surveillances in Table 4.1-1 (items 12, 26, 
and 27), Table 4.1-2 (item 2), and Table 4.1-3 (item 12) will not 
involve a significant increase [in] the probability or consequences 
of an accident since the relocation of the Technical Specifications 
to administrative controls governed by 10 CFR 50.59 does not affect 
the availability or function of charging and boric acid flow paths. 
CVCS is not an initiator of an accident (the dilution event is 
equipment malfunction that is manually terminated) and the proposed 
change does not alter overall system operation, physical design, 
system configuration, or operational setpoints. There will be no 
significant increase in the consequences of an accident because the 
required boration flow paths will continue to be available for 
boration to the reactor coolant system.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response: Relocation (i.e., removal from TS) of TS 3.2, the 
bases and the associated surveillances in Table 4.1-1 (items 12, 26, 
and 27), Table 4.1-2 (item 2), and Table 4.1-3 (item 12) will not 
create the possibility of a new or different kind of accident from 
any previously evaluated since it does not alter the overall system 
operation, physical design, system configuration, or operational 
setpoints. The plant systems for boration are operated in the same 
manner as before and, consequently, the relocation does not 
introduce any new accident initiators or failure mechanisms and does 
not invalidate the existing dilution event response. The boration 
function is not an accident initiator.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: Relocation (i.e., removal from TS) of TS 3.2, the 
bases and the associated surveillances in Table 4.1-1 (items 12, 26, 
and 27), Table 4.1-2 (item 2), and Table 4.1-3 (item 12) will not 
involve a significant reduction in margin of safety. The relocation 
is a change to the administrative controls that are used to assure 
system availability and those administrative controls are governed 
by 10 CFR 50.59. The manner in which the system is operated does not 
change and there is no change to physical design, system 
configuration, or operational setpoints. Previous analyses of system 
malfunction remain unchanged. The current Technical Specification 
does not meet the criteria in 10 CFR 50.36(c)(2)(ii) for inclusion 
in the technical specifications.

    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 10601.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Director.

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

    Date of amendment request: December 30, 1998.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Limiting Condition for Operation 
(LCO) 3.7.3 and Table 3.7.3-1. The proposed changes would modify the 
flood protection actions required during periods of elevated river 
water level.
    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 changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed TS revisions related to flood protection TS Action 
Statements involve no hardware changes and no changes to existing 
structures, systems or components. The proposed changes to the flood 
protection TS Action Statements ensure that the supported systems 
can perform their required safety functions under worst case design 
basis conditions, consistent with limitations imposed by other TS. 
The proposed flood protection TS ACTION Statements ensure that the 
plant is directed to enter a safe shutdown condition whenever the 
capability to withstand worst case design basis conditions is 
affected. Since the flood protection changes will still ensure that 
the plant remains capable of meeting applicable design basis 
requirements and retains the capability to mitigate the consequences 
of accidents described in the [Hope Creek] HC [Updated Final Safety 
Analysis Report] UFSAR, the proposed changes were determined to be 
acceptable. As a result, these changes will neither increase the 
probability of an accident previously evaluated nor increase the 
radiological dose 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 to the flood protection TS contained in 
this submittal will not adversely impact the operation of any safety 
related component or equipment. Since the proposed changes involve 
no hardware changes and no changes to existing structures, systems 
or components, there can be no impact on the potential occurrence of 
any accident due to new equipment failure modes. The resulting 
operational limits imposed by the flood protection LCO ensure that 
the plant can either perform its design basis safety functions or an 
appropriately conservative shutdown action statement is entered. 
Furthermore, there is no change in plant testing proposed in this 
change request that could initiate an event. Therefore, these 
changes will 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 changes for the flood protection TS retain the 
plant's continued capability to withstand worst case design basis 
conditions. The proposed flood protection TS ACTION Statements 
ensure that the plant is directed to: (1) enter a safe shutdown 
condition whenever the capability to withstand worst case design 
basis conditions is lost; or (2) enter a conservatively short period 
of continued operation when supported system redundancy is reduced. 
Since the plant will still remain capable of meeting all applicable 
design basis requirements and retaining the

[[Page 9201]]

capability to withstand worst case design basis events described in 
the HC UFSAR, the proposed changes were determined to 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: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: William M. Dean.

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

    Description of amendment request: The proposed changes revise the 
descriptive details of Technical Specification 4.7.1.2.1.a, regarding 
performance testing of the Auxiliary Feedwater (AFW) pumps, to more 
closely adhere to NUREG-1431, Improved Standard Technical 
Specifications for Westinghouse Plants. This involves relocating the 
surveillance-required numerical values for the AFW pump performance 
test discharge pressure and flow rate to the South Texas Project 
Updated Final Safety Analysis Report (UFSAR).
    Date of amendment request: January 20, 1999.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change, which relocates descriptive details (i.e., 
numerical values for AFW pump discharge pressure and flow rate) of 
the surveillance testing applicable to the AFW pumps, does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated. The affected AFW pump testing 
pressure and flow descriptive details that are being removed from 
SRs 4.7.1.2.1.a.1 and 4.7.1.2.1.a.2 are not related to any assumed 
initiators of analyzed events and are not assumed to mitigate 
accident or transient events. The requirement to perform testing on 
a monthly, staggered basis is not altered by the proposed change, 
and will remain in the Technical Specifications. The descriptive 
details of the surveillance testing will be relocated from the 
Technical Specifications to the USFAR and will be maintained 
pursuant to 10CFR50.59. The proposed revised wording of SRs 
4.7.1.2.1.a.1 and 4.7.1.2.1.a.2 (i.e., to verify the developed head 
of each pump is greater than or equal to the required developed 
head) and the relocation of pump testing details to the UFSAR is 
consistent with the AFW pump test requirements in NUREG-1431. In 
addition, the surveillance testing details are addressed in existing 
surveillance procedures that are also controlled by 10CFR50.59 and 
subject to the change control provisions imposed by plant 
administrative procedures, which endorse applicable regulations and 
standards. Therefore, this proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change relocates descriptive details (i.e., 
numerical values for AFW pump discharge pressure and flow rate) of 
surveillance testing applicable to the AFW pumps, which do not meet 
the criteria for inclusion in Technical Specifications as identified 
in 10CFR50.36(c)(3). The requirement to perform testing on a 
monthly, staggered basis is not altered by the proposed change, and 
will remain in the Technical Specifications. Additionally, 
relocation of the descriptive testing details is consistent with the 
wording of the AFW pump test requirements in NUREG-1431, which does 
not specify minimum numerical pressure and flow limits. The proposed 
change does not involve a physical alteration of the plant (no new 
or different type of equipment will be installed) or make changes in 
the methods governing normal plant operation. The change will not 
impose different requirements, and any future changes to these 
relocated surveillance testing details or to the applicable 
surveillance procedures will be evaluated per the requirements of 
10CFR50.59. This change will not alter assumptions made in the 
safety analysis and licensing basis. Therefore, this 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 a margin 
of safety?
    The proposed change, which relocates descriptive details (i.e., 
numerical values for AFW pump discharge pressure and flow rate) of 
the surveillance testing applicable to the AFW pumps, will not 
reduce a margin of safety since it has no impact on any safety 
analysis assumptions. The requirement to perform AFW pump testing on 
a monthly, staggered basis will not be altered by the proposed 
change, and will remain in the Technical Specifications. 
Furthermore, the proposed change will not affect the operability 
requirements of the AFW system as delineated in Specification 
3.7.1.2. Since any future changes to these relocated surveillance 
testing details or to the applicable surveillance procedures will be 
evaluated per the requirements of 10CFR50.59, there is no reduction 
in a margin of safety. Finally, this proposed change is also 
consistent with NUREG-1431, previously approved by the NRC Staff. 
Revising the Technical Specifications to reflect the approved NUREG-
1431 content ensures no significant reduction in the margin of 
safety. Therefore, this proposed change does not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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, TX 77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
    NRC Project Director: John N. Hannon.

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

    Date of application for amendments: January 15, 1999 (TS 98-07).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah (SQN) Technical Specification (TS) requirements by 
adding a new action statement to TS 3.1.3.2, ``Position Indicating 
Systems--Operating,'' that eliminates the need to enter TS 3.0.3 
whenever two or more individual rod position indicators (RPIs) may be 
inoperable per bank, while maintaining the appropriate overall level of 
protection and adding flexibility to the initial determination of the 
position of the non-indicating rod(s).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change to TS 3.1.3.2 does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The potential for the new action statement to 
impact the probability or consequences of the safety analyses for 
the plant lies only in the area of operator-exacerbated reactivity 
events due to

[[Page 9202]]

a loss of RCCA [rod control cluster assembly] position indication.
    RCCA events such as: One or more dropped RCCAs, a dropped RCCA 
bank or a RCCA ejection (FSAR [Final Safety Analysis Report] 
Sections 15.2.3 and 15.4.6, respectively) are not impacted since the 
new action statement does not involve a design change. Events such 
as: Uncontrolled RCCA bank withdrawal at power, statically 
misaligned RCCA or withdrawal of a single RCCA (FSAR Sections 
15.2.2, 15.2.3, and 15.3.6, respectively) involve, or potentially 
involve, operator action and are of interest. The uncontrolled RCCA 
bank withdrawal at power is an ANS [American Nuclear Society] 
Condition II transient that has been analyzed using a positive 
reactivity insertion rate greater than that for the simultaneous 
withdrawal of the two control banks having the maximum combined 
worth at maximum speed. Whether the event is caused by a failure in 
the rod control system or by operator error has no effect on the 
positive reactivity insertion rate assumed in the analysis. The 
protection systems assumed in the analysis are unaffected since 
there is no change to the design. Loss of the RPIS would not result 
in more frequent control rod movement by plant operators. Therefore, 
the new action statement would not affect the analysis of this event 
and departure from nucleate boiling ratio (DNBR) design basis would 
still be met.
    The most severe misalignment situation, with respect to DNBR, 
arises from cases in which one RCCA is fully inserted or where Bank 
D is fully inserted to its insertion limits with one RCCA fully 
withdrawn. For these cases, as discussed in FSAR Section 15.2.3.2, 
the DNBR remains above the safety analysis limit values. Also, the 
control bank insertion limit alarms remain available to warn 
operators that bank insertion limits have been reached.
    A compensatory action associated with this new action statement, 
placing the control rods under manual control, addresses concerns 
associated with automatic rod motion due to the rod control system 
and inadvertent operator contribution to these events.
    The worst-case event of those described above, the withdrawal of 
a single RCCA, is an ANS Condition III event. It has been analyzed 
in FSAR Section 15.3.6, assuming that operators ignore RCCA position 
indication or that multiple rod control system failures occur. No 
single electrical or mechanical failure in the rod control system 
could cause the accidental withdrawal of a single RCCA from an 
inserted bank at full power operation. The operator could 
deliberately withdraw a single RCCA in the control bank. This 
feature is necessary in order to retrieve an accidentally dropped 
rod. This new action statement does not change the plant design; 
therefore, there would be no change in the probability of the event 
being induced by the unlikely, simultaneous electrical failures 
(FSAR Section 7.7.2.2).
    The change in the time to determine the position of the non-
indicating rods, indirectly with the movable incore detectors, does 
not involve a design change nor does it affect the immediate 
response of the operator to the event, therefore, it does not affect 
the results of the analyses described above.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Since there is no change to the design associated with the 
proposed change, it does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed change involves a loss of the RPIS [Rod Position 
Indication System] and establishes compensatory measures to maintain 
control rod position consistent with the assumptions used in the 
existing accident and transient analyses. The new action statement 
provides sufficient time for troubleshooting while avoiding 
unnecessary plant shutdowns per TS 3.0.3.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change to TS 3.1.3.2 does not involve a significant 
reduction in a margin of safety. As discussed in Section IV.A above, 
the results of the FSAR Chapter 15 safety analyses for the 
applicable events, are not affected by the proposed changes. 
Therefore, the safety margins demonstrated by these analyses remain 
unchanged. The additional time to obtain the flux maps is consistent 
with the 12-hour time frame allowed to verify shutdown margin when a 
rod is misaligned from its group step counter height by more than 
plus or minus 12 steps in TS 3.1.3.1 and remains within a shiftly 
basis. Therefore, it does not reduce the margin of safety.

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

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

    Date of amendment request: January 29, 1999 (TSCR 211).
    Description of amendment request: The proposed amendments reflect 
changes to sections 15.6 and 15.7 of the Point Beach Nuclear Plant 
(PBNP), Units 1 and 2, Technical Specifications (TS). The proposed 
changes are considered administrative in nature and reflect personnel 
title changes, an increase in minimum operating crew shift staffing, 
relocation of the Manager's Supervisory Staff composition and 
functional requirements to owner controlled documents, and revisions to 
the procedure review and approval process.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not result in a significant increase in 
the probability or consequences of an accident previously evaluated.
    These changes are administrative and therefore do not involve a 
significant increase in the probability of an accident previously 
evaluated because no such accidents are affected by the proposed 
revisions. The proposed TS changes do not introduce any new accident 
initiators since no accidents previously evaluated have as their 
initiators anything related to the administrative changes described 
above.
    In addition, initiating conditions and assumptions are unchanged 
and remain as previously analyzed for accidents in the PBNP Final 
Safety Analysis Report. The proposed TS changes do not involve any 
physical changes to systems or components, nor do they alter the 
typical manner in which the systems or components are operated. All 
Limiting Conditions [for] Operation, Limiting Safety System 
Settings, and Safety Limits specified in the TS remain unchanged. 
Therefore, these changes do not increase the probability of 
previously evaluated accidents.
    These changes do not involve a significant increase in the 
consequences of an accident previously evaluated because the source 
term, containment isolation or radiological releases are not being 
changed by these proposed revisions. Existing system and component 
redundancy and operation is not being changed by these proposed 
changes. The assumptions used in evaluating the radiological 
consequences in the PBNP Final Safety Analysis Report are not 
invalidated; therefore, these changes do not affect the consequences 
of previously evaluated accidents.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    These changes do not introduce nor increase the number of 
failure mechanisms of a new or different type than those previously 
evaluated since there are no physical changes being made to the 
facility. The design and design basis of the facility remain 
unchanged. The plant safety analyses remain unchanged. Therefore, 
the possibility of a new or different kind of accident from any 
accident previously evaluated is not introduced.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not involve a significant reduction in 
a margin of safety.

[[Page 9203]]

    The proposed changes do not involve a significant reduction in 
the margin of safety because existing component redundancy is not 
being changed by these proposed changes. There are no new or 
significant changes to the initial conditions contributing to 
accident severity or consequences, and safety margins established 
through the design and facility license including the Technical 
Specifications remain unchanged. Therefore, there are no significant 
reductions in a margin of safety introduced by [these] proposed 
amendment[s].

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

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

    Date of amendment request: December 29, 1998.
    Description of amendment request: This amendment would revise the 
Wolf Creek Technical Specification (TS) Figures 3.4-2, 3.4-3, and 3.4-4 
to incorporate revised reactor coolant system heatup and cooldown limit 
curves and a revised cold overpressure mitigation system (COMS) power 
operated relief valve (PORV) setpoint limit curve.
    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 [reactor coolant system] 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 Project Director: William H. Bateman.

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

    Date of amendment request: January 12, 1999.
    Description of amendment request: This license amendment request 
proposes to revise Wolf Creek Generating Station (WCGS) Technical 
Specification 3/4.7.5, Ultimate Heat Sink, to add a new action 
statement. Specifically, the new action statement will require 
verification of operability of the two residual heat removal (RHR) 
trains, or initiation of power reduction with only one RHR train 
operable, when the plant inlet water temperature is between 90 and 94 
degrees Farenheit. The current TS requires shutdown when plant inlet 
water temperature exceeds 90 degrees Farenheit.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not involve any physical alteration of 
plant systems, structures or components. The proposed change 
provides an allowed time for the plant to continue operation with 
plant inlet water temperature in excess of the current technical 
specification limit of 90 degrees Fahrenheit, up to 94 degrees 
Fahrenheit, which is less than the design limit of 95 degrees 
Fahrenheit for plant components. The plant inlet water temperature 
is not assumed to be an initiating condition of any accident 
analysis evaluated in the updated safety analysis report (USAR). 
Therefore, the allowance of a limited time for the water temperature 
to be in excess of the current limit does not involve an increase in 
the probability of an accident previously evaluated in the USAR. The 
UHS [ultimate heat sink] supports operability of safety related 
systems used to mitigate the consequences of an accident. Plant 
operation for brief periods with plant inlet water temperature 
greater than 90 degrees Fahrenheit up to 94 degrees Fahrenheit will 
not adversely affect the operability of these safety-related systems 
and will not adversely impact the ability of these systems to 
perform their safety-related functions. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated in the USAR.
    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 does not involve any physical alteration of 
plant systems,

[[Page 9204]]

structures or components. The temperature of the plant inlet water 
being greater than 90 degrees Fahrenheit but less than or equal to 
94 degrees Fahrenheit for a short period does not introduce new 
failure mechanisms for systems, structures or components not already 
considered in the USAR. Therefore, the possibility of a new or 
different kind of accident from any accident previously evaluated is 
not created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change will allow an increase in plant inlet water 
temperature above the current technical specification limit of 90 
degrees Fahrenheit for the Ultimate Heat Sink, and delay the 
requirement to shutdown the plant when the plant inlet water system 
temperature limit is exceeded for 12 hours. The proposed change does 
not alter any safety limits, limiting safety system settings, or 
limiting conditions for operation, and the proposed temperature 
increase will remain below the design limit cooling water input 
value for safety-related equipment, except for the unlikely event of 
a combination of a worst dam failure occurring with a loss of 
coolant accident during a period of severe meteorological 
conditions. Thus, the proposed change does not involve a significant 
reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    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 Project Director: William H. Bateman.

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

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

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa

    Date of application for amendment: January 22, 1999.
    Brief description of amendment: The amendment would revise 
Technical Specification Surveillance Requirement 3.8.1.7 to better 
match plant conditions during diesel generator (DG) testing by 
clarifying which voltage and frequency limits are applicable during the 
transient and steady state portions of the DG start.
    Date of publication of individual notice in Federal Register: 
February 1, 1999 (64 FR 4902).
    Expiration date of individual notice: March 3, 1999.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, IA 52401.

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

    Date of application for amendment: January 20, 1999.
    Brief description of amendment request: The proposed amendment 
requests changes to the Technical Specification degraded voltage relay 
setpoints.
    Date of publication of individual notice in Federal Register: 
January 28, 1999 (64 FR 4474).
    Expiration date of individual notice: March 1, 1999.
    Local Public Document Room location: Vespasian Warner Public 
Library, 310 N. Quincy Street, Clinton, IL 61727.

PP&L, Inc., Docket No. 50-388, Susquehanna Steam Electric Station, Unit 
2, Luzerne County, Pennsylvania

    Date of amendment request: November 23, 1998.
    Brief description of amendment request: The requested changes would 
change the allowable values for both the core spray system and the low 
pressure coolant injection system reactor steam dome pressure-low 
functions.
    Date of publication of individual notice in Federal Register: 
February 1, 1999 (64 FR 4904).
    Expiration date of individual notice: March 3, 1999.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

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.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois; Docket 
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and 
2, Rock Island County, Illinois

    Date of application for amendments: November 30, 1998, as 
supplemented by letter dated January 8, 1999.
    Brief description of amendments: The amendments relocate the 
requirement for removal of the Reactor Protection System (RPS) shorting 
links to the Updated Final Safety Analysis Report (UFSAR).

[[Page 9205]]

    Date of issuance: February 8, 1999.
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 170; 165 & 183; 180.
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 7, 1999. (64 FR 
1032).
    The January 8, 1999, submittal provided additional clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021.

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

    Date of application for amendment: March 27, 1998 (NRC-98-0033).
    Brief description of amendment: The amendment revises technical 
specifications (TS) 3.5.2 and 3.5.3 and the associated Bases, raising 
the minimum water level for the core spray system in the condensate 
storage tank (CST). The amendment also removes incorrect information 
from TS 3.5.3 regarding water inventory in the CST reserved for the 
high pressure coolant injection and reactor core isolation cooling 
systems.
    Date of issuance: February 8, 1999.
    Effective date: February 8, 1999, with full implementation within 
90 days.
    Amendment No.: 131.
    Facility Operating License No. NPF-43. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19967).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
Ellis Reference and Information Center, 3700 South Custer Road, Monroe, 
Michigan 48161.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: March 25, 1998, as supplemented by 
letter dated November 30, 1998.
    Brief description of amendment: The amendment changes the Appendix 
A Technical Specifications (TSs) by modifying TS 3.9.8.1, ``Shutdown 
Cooling and Coolant Circulation-High water Level,'' and TS 3.9.8.2, 
``Shutdown Coolant Circulation-Low Water Level,'' to change the minimum 
water level above the fuel assemblies seated in the reactor vessel at 
which the Shutdown Cooling System (SDC) is required to be maintained 
operable, or be in operation. Also TS 3.8.1.2, ``Electric Power Systems 
A.C. Sources Shutdown,'' and appropriate Bases are revised to make 
wording consistent with the TS 3.9.8.1 and 3.9.8.2.
    Date of issuance: February 2, 1999.
    Effective date: This license amendment is effective as of its date 
of issuance, to be implemented within 60 days.
    Amendment No.: 148.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 6, 1998 (63 FR 
25109).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 2, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

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

    Date of application for amendment: October 27, 1998.
    Brief description of amendment: This amendment revises TS 3/
4.8.2.3, ``Electrical Power Systems--DC Distribution--Operating,'' and 
the associated bases. The surveillance requirements for battery testing 
have been revised.
    Date of issuance: February 9, 1999.
    Effective date: February 9, 1999.
    Amendment No.: 229.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 64125).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 9, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.

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

    Date of application for amendment: June 30, 1998, as supplemented 
on December 9, 1998.
    Brief description of amendment: This amendment revised Technical 
Specification 3.1.7, ``Standby Liquid Control System,'' by increasing 
the boron concentration in the Standby Liquid Control System for Cycle 
8 fuel design.
    Date of issuance: February 8, 1999.
    Effective date: February 8, 1999.
    Amendment No.: 97.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 29, 1998 (63 FR 
40562).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081.

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

    Date of application for amendments: October 27, 1998.
    Brief description of amendments: The amendments revised Turkey 
Point Units 3 and 4 Technical Specifications to add the qualifications 
for the multi-discipline supervisor.
    Date of issuance: February 3, 1999.
    Effective date: February 3, 1999.
    Amendment Nos.: 199 and 193.
    Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 16, 1998 (63 
FR 69341).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Florida International

[[Page 9206]]

University, University Park, Miami, Florida 33199.

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

    Date of application for amendment: July 21, 1998, as supplemented 
October 6, December 16, and December 31, 1998.
    Brief description of amendment: The amendment changes various 
Reactor Protection System (RPS) and Engineered Safety Feature Actuation 
System setpoints and allowable values; corrects the specified maximum 
reactor power level limited by the high power level RPS trip; adds a 
new Technical Specification associated with the automatic isolation of 
steam generator blowdown; and makes several editorial changes to 
correct various errors and to provide needed clarification. The 
amendment also makes changes to the applicable Bases pages and expands 
the Bases to discuss the new requirements for the automatic isolation 
of steam generator blowdown. However, the staff has not completed its 
evaluation of the requested change in the trip setpoint and allowable 
values for the steam generator water level. This portion of the request 
will be addressed later.
    Date of issuance: February 8, 1999.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 226.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 12, 1998 (63 FR 
43208).
    The October 6, December 16, and December 31, 1998, letters provided 
clarifying information that did not change the scope of the July 21, 
1998, application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

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

    Date of application for amendment: March 3, 1998, as supplemented 
May 7, 1998.
    Brief description of amendment: The amendment revises the Millstone 
Unit 3 licensing basis by eliminating the requirement to have the 
recirculation spray system directly inject into the reactor coolant 
system following a design-basis accident, with the exception of loss-
of-coolant accident (LOCA) scenarios involving a long-term passive 
failure. The Millstone Unit 3 licensing basis maintains the direct 
injection requirement for scenarios, as a contingency, for situations 
where it may be needed--as in the case of a LOCA with a long-term 
passive failure or for beyond design-basis scenarios.
    Date of issuance: January 20, 1999.
    Effective date: As of the date of issuance, to be implemented 
within 60 days from the date of issuance.
    Amendment No.: 165.
    Facility Operating License No. NPF-49: Amendment revised the 
Millstone Unit 3 licensing basis.
    Date of initial notice in Federal Register: March 25, 1998 (63 FR 
14487).
    The May 7, 1998, letter provided clarifying information that did 
not change the scope of the March 3, 1998, application and the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment and final no 
significant hazards consideration determination are contained in a 
Safety Evaluation dated January 20, 1999.
    No significant hazards consideration comments received: No public 
comments received.
    A petition to intervene was received from the Citizens Regulatory 
Commission that was dismissed and terminated by the NRC Atomic Safety 
Licensing Board.
    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.

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

    Date of application for amendment: April 16, 1998.
    Brief description of amendment: The amendment changes the Technical 
Specifications to modify a testing requirement for the emergency diesel 
generators.
    Date of issuance: February 9, 1999.
    Effective date: February 9, 1999.
    Amendment No.: 187.
    Facility Operating License No. DPR-64: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 21, 1998, (63 
FR 56256).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 9, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.

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

    Date of application for amendment: June 16, 1998.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Section 6 to relocate the Safety Review Committee 
Reviews, Audits and Records from TS to the Quality Assurance Program 
Section of the Final Safety Analysis Report.
    Date of issuance: February 8, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 251.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 15, 1998 (63 FR 
38204).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: October 19, 1998.
    Brief description of amendment: This amendment eliminates 
restrictions imposed by Technical Specification (TS) 3.0.4 for the 
Filtration, Recirculation and Ventilation System

[[Page 9207]]

during fuel movement and CORE ALTERATION activities. Specifically, TS 
Limiting Conditions for Operation 3.6.5.3.1 and 3.6.5.3.2 have been 
revised to add a note stating that the provisions of TS 3.0.4 are not 
applicable for initiation of handling of irradiated fuel in the 
secondary containment and CORE ALTERATIONS provided that the plant is 
in OPERATIONAL CONDITION 5, with reactor water level equal to or 
greater than 22 feet 2 inches.
    Date of issuance: February 4, 1999.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 113.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 4121).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 4, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.

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

    Date of application for amendment: September 8, 1998, as 
supplemented December 8, 1998.
    Brief description of amendment: This amendment revised Appendix C, 
``Additional Conditions,'' and will allow the performance of single 
cell charging and the use of non-Class 1E single cell battery chargers, 
with proper electrical isolation, for charging connected cells in 
OPERABLE Class 1E batteries. The single cell chargers will be used to 
restore individual cell parameters to the normal limits specified in 
Technical Specification Table 4.8.2.1-1.
    Date of issuance: February 9, 1999.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 114.
    Facility Operating License No. NPF-57: This amendment revised 
Appendix C of the license.
    Date of initial notice in Federal Register: October 7, 1998 (63 FR 
53954).
    The December 8, 1998, supplement provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination or expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 9, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.

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

    Date of application for amendment: April 28, 1998, as supplemented 
September 29, 1998, and December 8, 1998.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.4.2.1 to replace the 1% setpoint 
tolerance limit for safety/relief valves (SRVs) with a 3% 
setpoint tolerance limit. In addition, the amendment revises TS 4.4.2.2 
to state that all SRVs will be re-certified to meet a 1% 
tolerance prior to returning the valves to service after setpoint 
testing.
    Date of issuance: February 10, 1999.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 115.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 17, 1998 (63 FR 
33108).
    The September 29, 1998, and December 8, 1998, supplements provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination or expand the scope of 
the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 10, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.

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

    Date of application for amendment: June 25, 1998, as supplemented 
August 25, 1998, and December 15, 1998.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) Surveillance Requirement 4.5.1.d.2.b by deleting the 
requirement to perform in-situ functional testing of the Automatic 
Depressurization System safety relief valves (SRVs) during startup 
testing activities. The amendment also revised TS Surveillance 
Requirement 4.4.2.1 such that the 18-month channel calibration for the 
SRV acoustic monitors will no longer require an exception to the 
provisions of TS 4.0.4, nor adjustments to SRV full open noise levels.
    Date of issuance: February 10, 1999.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 116.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 12, 1998 (63 FR 
43212).
    The August 25, 1998, and December 15, 1998, supplements provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination or expand the scope of 
the original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 10, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.

Public Service Electric & Gas Company, Docket No. 50-311, Salem Nuclear 
Generating Station, Unit No. 2, Salem County, New Jersey

    Date of application for amendment: October 12, 1998.
    Brief description of amendment: This amendment allowed a one-time 
extension of the Technical Specification (TS) surveillance interval to 
the end of fuel Cycle 10 for certain TS surveillance requirements 
(SRs). Specifically, the amendment extended the surveillance interval 
in (a) SR 4.3.2.1.3 for the instrumentation response time testing of 
each engineered safety features actuation system function, (b) SRs 
4.8.2.3.2.f and 4.8.2.5.2.d for service testing of the 125-volt DC and 
the 28-volt DC distribution system batteries, respectively, and (c) SR 
4.8.2.5.2.c.2 for verification that the 125-volt DC battery connections 
are clean, tight, and coated with anti-corrosion material. Because of 
the length of the last outage and delays in restart, the SRs would have 
become overdue prior to reaching the next refueling outage (2R10). The 
SRs are to be completed during the 2R10 outage, prior to returning the 
unit to Mode 4 (hot shutdown) upon outage completion.
    Date of issuance: February 1, 1999.

[[Page 9208]]

    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment No.: 198.
    Facility Operating License No. DPR-75: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 4, 1998 (63 FR 
59594).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 1, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

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

    Date of application for amendments: November 14, 1997, as 
supplemented by letters dated March 13, 1998, and November 10, 1998.
    Brief description of amendments: The amendments would revise the 
licensing basis as described in Section 3.5, ``Missile Protection,'' of 
the Updated Final Safety Analysis Report to allow the use of NUREG-
0800, ``Standard Review Plan'' methodology in evaluating tornado-
generated missiles.
    Date of issuance: February 9, 1999.
    Effective date: February 9, 1999, to be implemented in the next 
periodic update of the Updated Final Safety Analysis Report (UFSAR) in 
accordance with 10 CFR 50.71(e) that occurs after 60 days of the date 
of issuance.
    Amendment Nos.: Unit 2--148; Unit 3--140.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the UFSAR.
    Date of initial notice in Federal Register: December 31, 1997 (62 
FR 68315).
    The March 13, 1998, and November 10, 1998, supplemental letters 
provided additional clarifying information and did not change the 
original no significant hazards consideration determination. The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated February 9, 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., et al. Docket Nos. 50-424 and 
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of application for amendments: June 26, 1998, as supplemented 
by letters dated September 18 and November 30, 1998.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) as follows: (1) The Applicability of 
Limiting Condition for Operation (LCO) 3.3.6, ``Containment Ventilation 
Isolation Instrumentation,'' is revised to refer to TS Table 3.3.6-1; 
the TS table is revised to add a column entitled ``Applicable Modes or 
Other Specified Conditions.'' Then, the applicable modes for Manual 
Initiation, Automatic Actuation Logic and Actuation Relays, and Safety 
Injection are revised to include only Modes 1, 2, 3, and 4. Consistent 
with this change, LCO 3.3.6, Condition C and Required Action C.2 are 
revised to reflect that system level manual initiation and automatic 
actuation are not required during core alterations and/or during 
movement of irradiated fuel assemblies within containment. Appropriate 
Bases changes are included to reflect the TS changes. (2) LCO 3.9.4 is 
revised to allow the emergency air lock to be open during core 
alterations and/or during movement of irradiated fuel assemblies within 
containment. In addition, the LCO statement is revised to reflect that 
containment ventilation isolation (CVI) would be accomplished by 
manually closing the individual containment purge supply and exhaust 
isolation valves as opposed to a system level manual or automatic 
initiation, consistent with the proposed change to LCO 3.3.6. 
Surveillance Requirement (SR) 3.9.4.2 is revised to reflect the change 
to CVI. Appropriate Bases changes are included to reflect the TS 
changes. (3) LCO 3.7.6 is revised to delete the words ``Redundant 
CSTs'' from the title and LCO 3.7.6a is deleted. Appropriate Bases 
changes are included to reflect the changes.
    Date of issuance: January 29, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: Unit 1--105; Unit 2--83.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: October 7, 1998 (63 FR 
53955). The supplement dated November 30, 1998, provided clarifying 
information that did not change the scope of the application and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 29, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia.

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

    Date of application for amendments: July 13, 1998, as supplemented 
by letters dated December 16, 1998, and January 13, 1999.
    Brief description of amendments: The amendments revise Technical 
Specification Section 1.1, Definitions, for ``Engineered Safety Feature 
[ESF] Response Time'' and ``Reactor Trip System [RTS] Response Time'' 
to provide for verification of response time for selected components 
provided that the components and the methodology for verification have 
been previously reviewed and approved by the NRC.
    Date of issuance: February 8, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: 106 and 84.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 7, 1998 (63 FR 
53957).
    The December 16, 1998, and January 13, 1999, letters provided 
clarifying information that did not change the scope of the July 13, 
1998, application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia.

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: October 29, 1998.
    Brief description of amendments: Relocates portions of Technical 
Specification 4.8.1.1.2.g requirements regarding maintenance of the 
diesel generator fuel oil storage tank to the Technical Requirements 
Manual.

[[Page 9209]]

    Date of issuance: February 8, 1999.
    Effective date: The license amendment is effective as of its date 
of issuance, to be implemented within 30 days of issuance.
    Amendment Nos.: Unit 1--Amendment No. 102; Unit 2--Amendment No. 
89.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 16, 1998 (63 
FR 69347).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

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

    Date of application for amendments: November 16, 1998.
    Brief description of amendments: The amendments revise the Sequoyah 
Nuclear Plant Technical Specification (TS) emergency diesel generator 
surveillance requirements. The U.S. Nuclear Regulatory Commission staff 
has found the proposed changes to be acceptable.
    Date of issuance: February 9, 1999.
    Effective date: As of the date of issuance to be implemented no 
later than 45 days after issuance.
    Amendment Nos.: 242 and 232.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TSs.
    Date of initial notice in Federal Register: December 2, 1998 (63 FR 
66603).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 9, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

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

    Date of application for amendment: August 20, 1998.
    Brief description of amendment: Revises Technical Specifications 
(TS) through deletion of definition of SITE BOUNDARY, moves site map 
from TS to Final Safety Analysis Report and deletion of an uneeded 
reference to the site map.
    Date of issuance: February 3, 1999.
    Effective date: February 3, 1999.
    Amendment No.: 150.
    Possession Only License No. DPR-3: Amendment revised the Technical 
Specifications.
    Date of initial notice in Federal Register: October 7, 1998 (63 FR 
53962). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Greenfield Community College, 
1 College Drive, Greenfield, Massachusetts 01301.

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

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 99-4391 Filed 2-23-99; 8:45 am]
BILLING CODE 7590-01-P