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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is publishing this regular biweekly

notice. Pub. L. 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 April 25, 1998, through May 8, 1998. The

last biweekly notice was published on May 6, 1998 (63 FR 25101).

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 June 19, 1998, 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.

[[Page 27758]]

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses.

If a hearing is requested, the Commission will make a final

determination on the issue of no significant hazards consideration. The

final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

request for a hearing. Any hearing held would take place after issuance

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555-0001, Attention: Rulemakings and

Adjudications Staff, or may be delivered to the Commission's Public

Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,

by the above date. A copy of the petition should also be sent to the

Office of the General Counsel, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for a hearing will

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

petition and/or request should be granted based upon a balancing of

factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this action, see the

application for amendment which is available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the local public document room for

the particular facility involved.

Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,

Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of amendment request: January 14, 1998.

Description of amendment request: The proposed amendments would

change the Technical Specifications to allow replacement of the 125

volt direct current (DC) AT&T batteries with new Charter Power Systems,

Inc. (C&D) batteries, and revise the crosstie loading limitation.

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:

A. The proposed change does not involve a significant increase

in the probability of consequences of an accident previously

evaluated.

The replacement C&D battery has been selected to meet or exceed

the design, functional, and operational requirements of those of the

present AT&T battery, including crosstie load limitations. The C&D

batteries are similar in design to the previously installed Gould

batteries (e.g. electrolyte specific gravity and construction of the

plates) except for capacity. The replacement C&D batteries have a

significantly larger capacity than either the previously installed

Gould, or the currently installed AT&T, batteries. This increased

capacity can provide additional margin for future use. Also, the C&D

batteries are qualified for a 20 year life and meet the latest

applicable standards. The short circuit current provided by the C&D

batteries is well within the interrupting capability of the existing

DC system [c]ircuit breakers.

Additionally, the crosstie limit is increased to take advantage

of the larger C&D battery capacity. The C&D batteries were sized

based on having sufficient capacity to energize the design basis DC

loads of an operating unit with the [Institute of Electrical and

Electronics Engineers] IEEE-485 design margin while maintaining the

desired limited DC load of 200 amps for a shutdown unit. This

proposed change allows use of the C&D batteries' larger capacity.

Also, although adherence to the performance testing intervals

stated in IEEE Std 450 could result in a planned shutdown and

possible subsequent increase in the probability of occurrence of an

accident (e.g. Turbine Trip), it would be part of a controlled and

planned shutdown, therefore the increases would not be considered

significant.

The overall design, function, and operation of the DC system and

equipment has not been altered by these changes. The proposed

changes do not affect any accident initiators of 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

[Updated Final Safety Analysis Report] Chapter 15. Therefore, there

is no increase in the probability or consequences of an accident

previously evaluated.

B. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The replacement C&D batteries will provide the same function as

those of the installed AT&T batteries and will be operated with the

same types of operational controls. These limits include battery

float terminal voltage, individual cell voltage and electrolyte

specific gravity, and crosstie loading. Crosstie conditions are

allowed under the present Technical Specifications. The crosstie

limit is increased to take advantage of the larger C&D battery

capacity. The remaining changes are administrative in nature or

provide clarification to maintain consistency with other Technical

Specifications.

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 create any new or common failure modes. 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.

C. The proposed change does not involve a significant reduction

in a margin of safety.

The replacement C&D batteries will meet or exceed the design,

functional, and qualifications of the installed AT&T batteries. The

proposed Technical Specification limitations for the C&D batteries

are derived from the same methodology as the AT&T batteries with

applied margins in accordance with IEEE 485. Increasing the crosstie

loading limits takes advantage of the larger C&D battery capacity

with its increased design margin. The proposed change to the

crosstie loading limit will continue to conservatively envelope the

postulated design requirements. The remaining changes are

administrative in nature or provide clarification to maintain

consistency with other 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.

Therefore, there is no reduction in the margin of safety.

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

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

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: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60603.

NRC Project Director: Stuart A. Richards.

[[Page 27759]]

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: May 27, 1997, as supplemented on August

1, 1997, and March 24, 1998.

Description of amendment request: The proposed amendments would

revise Technical Specification Section 6, ``Administrative Controls,''

to incorporate revised organizational titles and would delete the Unit

1 License Condition 2.C.(30)(a) related to the function of the Shift

Technical Advisor. In addition, the proposed amendments would change

the submittal frequency of the Radiological Effluent Release Report

from semiannually to annually. The proposed amendments will also make

several administrative and editorial changes. The staff's proposed no

significant hazards consideration determination for the requested

change was published on July 30, 1997 (62 FR 40848).

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:

A. The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed changes do not affect any accident initiators or

precursors and do not change or alter the design assumptions for

systems or components used to mitigate the consequences of an

accident. The proposed changes do not affect the design or operation

of any system, structure, or component in the plant. There are no

changes to parameters governing plant operation, and, no new or

different type of equipment will be installed.

The proposed changes provide clarification, consistency with

station procedures, programs, the Code of Federal Regulations (10

CFR), other Technical Specifications, and Improved Technical

Specifications. These changes do not impact any accident previously

evaluated in the UFSAR [Updated Final Safety Analysis Report]. There

is no relaxation of applicable administrative controls. Those

administrative requirements which have no effect on safe operation

of the plant are eliminated.

B. The proposed changes do not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed changes do not affect the design or operation of

any plant system, structure, or component. There are no changes to

parameters governing plant operation, and, no new or different type

of equipment will be installed.

C. The proposed changes do not involve a significant reduction

in a margin of safety.

The proposed changes do not affect the margin of safety for any

Technical Specification. The initial conditions and methodologies

used in the accident analyses remain unchanged; therefore, accident

analyses results are not impacted. Plant safety parameters or

setpoints are not affected. All responsibilities described in the

Technical Specifications for administrative controls will continue

to be performed by individuals possessing the requisite

qualifications. Clarifications, relocations, and nomenclature

changes neither result in a reduction of personnel responsibilities,

nor do they cause a relaxation of programmatic controls. There are

no resulting effects on plant safety parameters or setpoints.

Guidance has been provided in ``Final Procedures and Standards

on No Significant Hazards Considerations,'' Final Rule, 51 FR 7744,

for the application of standards to license change requests for

determination of the existence of significant hazards

considerations. This document provides examples of amendments which

are and are not considered likely to involve significant hazards

considerations. These proposed amendments most closely fit the

example of a purely administrative change to the Technical

Specifications to achieve consistency throughout the Technical

Specifications, correction of an error, or a change in nomenclature.

The proposed amendment does not involve a significant relaxation

of the criteria used to establish safety limits, a significant

relaxation of the bases for the limiting safety system settings, or

a significant relaxation of the bases for the limiting conditions

for operations. The proposed change does not reduce the margin of

safety as defined in the basis for any Technical Specification.

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

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

Attorney for licensee: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60603.

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: April 13, 1998.

Description of amendment request: Unreviewed Safety Question

involving additional manual actions incorporated in new fire protection

procedures as a result of a revised Appendix R Safe Shutdown Safety

Analysis.

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) No significant increase in the probability or consequences

of an accident previously evaluated is involved because of the

following:

Two types of previously evaluated accidents are relevant to this

criterion: (1) A fire; (2) other accidents evaluated in the Updated

Final Safety Analysis Report. For these previously evaluated

accidents, the change would not result in an increase in either

their probabilities of occurrence or the consequences of their

occurrence, for the following reasons:

The additional operator manual actions do not significantly

change the probability or consequences of a fire. The likelihood of

a fire is unchanged. Additional operations do not significantly

change the fire loading nor introduce significant new ignition

sources. The quantities and arrangement of combustible materials are

not changed through additional manual actions.

The consequences of a fire are unchanged because operator manual

actions serve to support the station's ability to achieve and

maintain shutdown in the event of a fire.

Additional manual operations are for purposes of safe shutdown

in the event of a fire in areas requiring alternate shutdown

capability and do not impact other accident scenarios. Also, there

is no increase in the predicted frequency of other accidents as a

result of this change. Accordingly there is no significant change in

the probability or consequences of other accidents previously

evaluated because they are independent of this change in procedures

for fire scenarios.

(2) The possibility of a new or different kind of accident from

any accident previously evaluated is not created because:

The proposed change does not create the possibility of a new or

different kind of accident from that previously evaluated for the

Quad Cities Station. Although the number of manual actions increased

and there may be some compression in the time for taking necessary

actions relative to the current safe shutdown analysis and

procedures, there is no significant change in the operation of plant

equipment following the postulated fire event. The existing safe

shutdown analysis already relies on operator manual actions which

perform the same type of actions.

The overall approach and methodology to performing these

operator actions are not significantly different from the prior

approach and methodology. This proposed change does not involve an

accident initiator or failure not previously considered. The results

or effects of equipment malfunctions

[[Page 27760]]

previously evaluated are unchanged as the result of potential

operator errors. No new failures would occur, and no new modes of

operation are introduced by the proposed changes.

Additional manual actions and the timing thereof provide a

somewhat different demand on the plant equipment operators, but

still provide an effective method for achieving and maintaining

post-fire safe shutdown for areas requiring alternate shutdown

capability. As such, the proposed changes do not create the

possibility of a new or different kind of accident.

(3) No significant reduction in the margin of safety is involved

because:

A change in the fire protection program does not result in a

significant reduction in the margin of safety if the change does not

result in a significant adverse impact on the plant's ability to

achieve and maintain safe shutdown in the event of a fire. The

proposed operator manual actions to achieve and maintain safe

shutdown in a fire scenario do not significantly affect the

capability or reliability of the equipment assumed to operate in the

safety analysis.

The types of manual actions to be performed in support of

Appendix R safe shutdown functions are not significantly different

from those previously considered. The complexity of actions is not

significantly changed. Indeed many of the additional actions are

designed to provide additional protection from spurious operations

which could result from a fire.

Any reduction in margin associated with changes in the time

before which certain manual actions must occur is largely a result

of re-analyses which incorporate conservatisms not previously

considered. In total, the proposed changes do not adversely impact

the capability to meet the requirements of Appendix R. Any reduction

in margin associated with additional manual actions to achieve and

maintain post fire safe shutdown in areas requiring alternate

capabilities does not involve a significant reduction in margin.

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 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: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60603.

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: May 27, 1997, as supplemented by letters

dated March 9, March 20, and April 20, 1998.

Description of amendment request: The proposed amendments would

revise the current Technical Specifications (TS) of each unit to

conform with NUREG-1431, Revision 1, ``Standard Technical

Specifications--Westinghouse Plants.'' The Commission had previously

issued a Notice of Consideration of Issuance of Amendments published in

the Federal Register on July 14, 1997 (62 FR 37628) covering all the

proposed changes that were indeed within the scope of NUREG-1431. In

DEC's March 9, March 20, and April 20, 1998, supplements, there are

proposed changes that are beyond the scope of NUREG-1431, which were,

thus, not covered by the staff's July 14, 1997, notice. The following

descriptions and proposed no significant hazard analyses cover only

those beyond-scope changes. Associated with each change are

administrative/editorial changes such that the new or revised

requirements would fit into the format of NUREG-1431.

1. Table 3.3-3 of the current TS contains an entry regarding the

Containment Pressure Control System, allowing an inoperable channel be

placed in trip in 1 hour. DEC proposed to tighten this requirement such

that the system supported by the inoperable channel be declared

inoperable immediately. No changes to the design of the Containment

Pressure Control System or other systems were proposed by DEC.

2. Table 4.3-1 of the Unit 1 current TS has a footnote (No. 13)

that specifies a filter time constant of 1.5 seconds in the steam

generator low-low level reactor trip circuitry. DEC proposed to delete

this time constant since it was never used. No design changes to the

instrumentation and control systems are involved.

3. Section 4.5.1.1.c of the current TS requires that power be

removed from the accumulator isolation valve when the reactor coolant

system pressure is greater than 2000 pounds per square inch gauge

(psig). DEC proposed to make this requirement more restrictive,

lowering this threshold to 1000 psig on the recommendation of the

nuclear vendor, Westinghouse. No design changes to the accumulator

system are involved.

4. Section 4.6.5.1.b.1 of the current TS requires that the boron

concentration of the ice in the ice condenser be verified once every 9

months to be at least 1800 ppm. DEC proposed to relax the frequency

from 9 months to 18 months on the basis that boron, in the form of

sodium tetraborate, does not decrease in quantity even though the ice

sublimates. No design changes to the ice condenser are involved.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), DEC has provided its

analyses of the issue of no significant hazards consideration for each

of the above proposed changes. The NRC staff has reviewed DEC's

analyses against the standards of 10 CFR 50.92(c). The NRC staff's

analysis is presented below.

1. Will the changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

For all the changes the answer is ``no.'' The proposed changes will

not affect the safety function of the subject systems. There will be no

direct effect on the design or operation of any plant structures,

systems, or components. No previously analyzed accidents were initiated

by the functions of these systems, and the systems were not factors in

the consequences of previously analyzed accidents. Therefore, the

proposed changes will have no impact on the consequences or

probabilities of any previously evaluated accidents.

2. Will the changes create the possibility of a new or difference

kind of accident from any accident previously evaluated? For all the

changes the answer is ``no.'' The proposed changes would not lead to

any hardware or operating procedure change. Hence, no new equipment

failure modes or accidents from those previously evaluated will be

created.

3. Will the changes involve a significant reduction in a margin of

safety?

For all the changes the answer is ``no.'' Margin of safety is

associated with confidence in the design and operation of the plant.

The proposed changes to the TS do not involve any change to plant

design, operation, or analysis. Thus, the margin of safety previously

analyzed and evaluated is maintained.

Based on this analysis, it appears that the three standards of 10

CFR 50.92(c) are satisfied for each of the proposed changes. 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.

[[Page 27761]]

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, et al., Docket Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: April 20, 1998.

Description of amendment request: The Control Room Area Ventilation

System (CRAVS) can be actuated by a number of ways, including by the

engineered safety features actuation signal (ESFAS) when safety

injection is also initiated. The only relationship between automatic

actuation of the CRAVS and the ESFAS is through safety injection

initiation, applicable in Modes 1, 2, 3, and 4. However, in Tables 3.3-

3 and 4.3-2 of the units' Technical Specifications, regarding

operability and surveillance requirements, the CRAVS automatic

actuation has been erroneously specified for all modes (Modes 1, 2, 3,

4, 5, and 6). The licensee proposed to correct this error by the

proposed amendment.

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. The Control Room Area Ventilation System and

ESFAS are not accident initiating systems; they are accident

mitigating systems. Therefore, changing the mode requirements for

the subject ESFAS functional unit cannot impact accident initiating

probabilities. The technical justification associated with this

proposed amendment shows that the current Technical Specification

mode requirements for the subject functional unit are incorrect as

written. The Control Room Area Ventilation System and ESFAS will

remain fully capable of performing their design accident mitigation

functions for the modes in which they are required. The Control Room

Area Ventilation System operability requirement of Technical

Specification 3/4.7.6 will continue to be met. Therefore, no

accident consequences will be impacted.

Second Standard

Implementation of this amendment would not create the

possibility of a new or different kind of accident from any accident

previously evaluated. As noted previously, the Control Room Area

Ventilation System and ESFAS are not accident initiating systems.

Correcting the mode requirements as specified will not impact any

plant systems that are accident initiators. No other modifications

are being proposed to the plant which would result in the creation

of new accident mechanisms. Also, no changes are being made to the

way in which the plant is operated, so no new failure mechanisms

will be initiated.

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 the fission

product barriers will not be impacted by implementation of this

proposed amendment. Both the Control Room Area Ventilation System

and the ESFAS will remain fully capable of performing their design

functions for the modes in which they are required. Therefore, no

safety margin will be significantly impacted.

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

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina. Attorney for licensee: 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 (DEC), Docket Nos. 50-369 and 50-370, McGuire

Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: May 27, 1997, as supplemented by letter

dated March 9, 1998.

Description of amendment request: The three proposed changes are

associated with DEC's application to convert to the Improved Technical

Specifications. The first change would increase the surveillance

interval for the boron concentration of the ice bed from once per 9

months, to every 18 months. This change is supported by operating

experience data, establishes surveillance intervals that coincide with

refueling outages, and minimizes containment entries during power

operation. The second change would decrease the Reactor Coolant System

pressure level at which power is removed from the accumulator isolation

valve from 2000 pounds per square inch gauge (psig) to 1000 psig. This

change is considered a more restrictive change, and is based on

recommendations by Westinghouse Nuclear Safety Advisory Letter 97-003.

The third change would revise the Turbine Trip and Feedwater Isolation

function to include an initiation signal from the average-low

temperature. This change is considered a more restrictive change, and

is consistent with the plant design and safety analysis.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration for each of the above proposed changes. The NRC staff has

reviewed the licensee's analyses against the standards of 10 CFR

50.92(c). The NRC staff's analysis is presented below:

1. Will the changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes will not affect the safety function of the

subject systems. There will be no direct effect on the design or

operation of any plant structures, systems, or components. No

previously analyzed accidents were initiated by functions of these

systems, and the systems were not factors in the consequences of

previously analyzed accidents. Therefore, the proposed changes will

have no impact on the consequences or probabilities of any previously

evaluated accidents.

2. Will the changes create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed changes would not lead to any hardware or operating

procedure change. Hence, no new equipment failure modes or accidents

from those previously evaluated will be created.

3. Will the changes involve a significant reduction in a margin of

safety?

Margin of safety is associated with confidence in the design and

operation of the plant. The proposed changes do not involve any change

to the plant design, operation, or analysis. Thus, the margin of safety

previously analyzed and evaluated is maintained.

Based on this analysis, it appears that the three standards of 10

CFR 50.92(c) are satisfied for each of the proposed changes. Therefore,

the NRC staff proposes to determine that the amendment request involves

no significant hazards consideration.

Local Public Document Room location: J. Murrey Atkins Library,

University of North Carolina at

[[Page 27762]]

Charlotte, 9201 University City Boulevard, North Carolina.

Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation,

422 South Church Street, Charlotte, North Carolina.

NRC Project Director: Herbert N. Berkow.

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: April 9, 1998.

Description of amendment request: The proposed amendment would

revise license condition 2.C(13) to allow Final Feedwater Temperature

Reduction (FFWTR) at the River Bend Station, Unit No.1(RBS). FFWTR is

to be used at the end of each fuel cycle to allow approximately

fourteen additional effective full power days of operation.

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 significantly increase the

probability or consequences of an accident previously evaluated.

The abnormal operational occurrences or accidents analyzed in

the SAR [Safety Analysis Report] have been examined for impact

caused by partial feedwater heating during cycle extension or at

coastdown condition. The limiting abnormal operation transients,

including the Load Rejection with no Bypass (LRNBP) event and the

Feedwater Controller Failure (FWCF) maximum demand event, Turbine

Trip with No Bypass (TTNBP) and Pressure Regulator Failure Downscale

(PRFD) have been analyzed based upon the core nuclear characteristic

at end-of-cycle (EOC) conditions including the effects of increased

core flow and the proposed reduction in feedwater temperature with

an all-rods-out condition.

The LOCA (Loss of Coolant Accident), fuel loading error, rod

drop accident, rod withdrawal error, overpressure protections and

ATWS (anticipated transient without scram) analyses have been

evaluated for the effects of reduced feedwater temperature operation

and found acceptable. In addition, the case of the analyzed

operational events the current fuel OLMCPR (operating limit minimum

critical power ratio) and MAPLHGR (maximum average planar linear

heat generation rate) limits bound those necessary for operation and

therefore, are not affected by operation with FFWTR therefore, these

events are bounded by the current RBS analysis. Because the accident

results are acceptable and the current operating fuel limits are

unaffected, the consequence of an event previously evaluated remains

unaffected.

The probability of an accident is not affected by the proposed

changes since no systems or equipment which could initiate an

accident are affected. Therefore, the proposed changes do not

significantly increase the probability or consequences of any

previously evaluated accident.

2. The request does not create the possibility of occurrence of

a new or different kind of accident from any accident previously

evaluated.

The FFWTR mode of operation is functionally similar to operation

with Feedwater Heaters Out of Service (USAR (Updated Safety Analysis

Report) Section 15.1.7). All abnormal operational transients or

accidents have been evaluated and the most limiting cases have been

analyzed for applicability for the FFWTR operation. Limits on

MAPLHGR and OLMCPR (including the power and flow dependent MCPR)

which are included in the Core Operating Limits Report (COLR) as

part of the normal reload licensing process will continue to assure

that operations are within the assumptions, initial conditions and

assumed power distribution and therefore will not create a new type

of accident. The proposed changes do not involve new setpoints, new

system interactions, or physical modifications to the plant.

Therefore, the proposed changes do not create the possibility of a

new or different kind of accident from any previous analyzed.

3. The request does not involve a significant reduction in a

margin of safety.

The proposed changes do not involve any setpoint changes and

would allow steady state power operation at off-rated feedwater

temperature conditions as defined in current plant procedures. The

transient and accidents described in the SAR are evaluated for

effects caused by the reduced feedwater temperature of 100 (degrees)

F. As described in Attachment 4 (to the April 9, 1998, amendment

request), * * * the FWCF is the most limiting transient under such

condition and the required OLMCPR for this event is bounded by the

EOC OLMCPR limits set forth in the RBS COLR. The thermal limits MCPR

and LHGR curves, and the MAPLHGR limits establish limits on power

operation and thereby ensure that the core is operated within the

assumptions and initial conditions of the transient or accident

analyses.

Operation within these limits set forth by the MCPR limits, the

LHGR limits and the MAPLHGR criteria will ensure that the margin of

safety will be maintained to the same level described in the

Technical Specifications Bases and the SAR. As a result the

consequences of postulated transients or accidents are not

increased.

The MCPR safety limit, mechanical performance limits and

overpressure limits are not exceeded during any transient or

postulated accident at normal feedwater temperature or at reduced

feedwater temperature condition. Therefore, the proposed changes to

allow partial feedwater heating for cycle extension do not involve a

significant reduction in 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, DC 20005.

NRC Project Director: John N. Hannon.

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

1, DeWitt County, Illinois

Date of amendment request: April 27, 1998.

Description of amendment request: The proposed amendment would

change the title of ``shift supervisor'' to ``shift manager'' in the

Technical Specifications (TS).

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 replaces the title of ``shift

supervisor'' with the title of ``shift manager'' as it pertains to

the responsibilities of the position described in TS Section 5.1.2.

The proposed change does not involve a change to the plant design or

to the operation of the plant by qualified operators and senior

operators. Although this change involves changes to the Operations

department, individuals in those positions comprising the operating

crews will continue to have to meet the same licensing, experience,

training, and education requirements, notwithstanding the proposed

change in the title of the individual with ultimate command

authority in the main control room, from ``shift supervisor'' to

``shift manager.'' Therefore, the operation of CPS is not affected

by this change. Further, as also noted, the proposed change does not

affect plant design. It therefore would not affect systems,

structures, or components important to safety, particularly those

associated with the plant accident analyses. As a result, the

proposed change does not affect any parameters or conditions that

may contribute to the initiation of any accidents previously

evaluated, nor does it affect the operation or response of systems,

structures, or components assumed to mitigate postulated accidents

that have been evaluated/analyzed. On this basis, IP has concluded

that the proposed change will not result in a significant increase

in the probability or consequences of any accident previously

evaluated.

(2) As noted above, the proposed change does not involve a

change to design or operation of the plant. As a result, the

proposed change, which is only administrative in nature, cannot

introduce

[[Page 27763]]

any new failure modes or precursors, parameters, or conditions that

could cause or contribute to the initiation of any new accidents not

previously evaluated. On this basis, IP has concluded that the

proposed change will not create the possibility of a new or

different kind of accident not previously evaluated.

(3) As noted above, the proposed change is an administrative

change that involves no changes to plant design or operation,

including the design or operation of systems, components, or

structures important to safety. On this basis there are no margins

of safety affected by the proposed change. As a result, IP has

concluded that the proposed change will not result in a significant

reduction in a margin of safety.

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

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, IL 61727.

Attorney for licensee: Leah Manning Stetzner, Vice President,

General Counsel, and Corporate Secretary, 500 South 27th Street,

Decatur, IL 62525.

NRC Project Director: Ronald R. Bellamy, Acting.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County, Maine

Date of amendment request: April 13, 1998.

Description of amendment request: The proposed amendment would

amend the Technical Specifications to base the Limiting Condition for

Operation for the fuel storage pool water level on a revised analysis

of the fuel handling accident and on a new analysis for radiological

shielding during movement of irradiated fuel.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change does not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The proposed restrictions on the water level in the spent fuel

pool has no impact on the probability or consequences of the

remaining applicable design basis accidents. These restrictions are

fulfilled by normal operating conditions, preserve initial

conditions assumed in the analyses of postulated DBAs and ensure

that the conditions of such DBAs are consistent with the analyses.

Revised analysis was performed assuming a fuel handling accident

occurs after the spent fuel fission products have decayed at least

1-year. The initial conditions assumed a minimum of 19 feet of water

for iodine absorption. No credit was taken for control room or spent

fuel pool ventilation filtration. The results of the revised

analysis demonstrate that the projected doses resulting from a

postulated fuel handling accident are insignificant in comparison to

10 CFR part 100 limits. Therefore, the proposed changes to the

Technical Specifications do not involve any increase in the

probability or consequences of any accident previously evaluated.

2. Create the possibility of a new or different kind of accident

from any accident previously evaluated.

The proposed restrictions on the water level in the spent fuel

pool are fulfilled by normal operating conditions and preserve

initial conditions assumed in the analysis of postulated DBAs. These

additional restrictions do not involve changes to any structure or

equipment affecting the safe storage of irradiated fuel. The results

of the revised analysis of a fuel handling accident demonstrate that

the projected doses are insignificant in comparison to 10 CFR part

100 limits with a minimum of 19 feet of water for iodine absorption.

In addition, maintaining this minimum water level will also provide

sufficient shielding for personnel radiation protection during fuel

movement. Therefore, the proposed changes to the Technical

Specifications would not create the possibility of a new or

different accident from any accident previously evaluated.

3. Involve a significant reduction in a margin of safety.

The proposed restrictions on the water level in the spent fuel

pool preserve initial conditions assumed in the analyses of

postulated DBAs and ensure that margins of safety contained in the

analyses are maintained. The margin of safety for the fuel handling

accident relates to the acceptance limit which the NRC approved

during its review of the license. The fuel handling accident

acceptance limit defined in the basis for the Maine Yankee Technical

Specification (formerly specified as TS 3.13.D.10) is 10% of 10 CFR

part 100 limits. A reduction in margin of safety occurs when the

acceptance limit would no longer be met as a result of a proposed

change. Since the acceptance limit is met, there is no reduction in

margin of safety. The projected dose rates at the specified Fuel

Storage Pool water level during fuel movement with a fuel assembly

raised to its highest allowable height would result in personnel

exposures within that previously assumed. There is no reduction in a

margin of safety. The NRC acceptance limit which is that combination

of occupancy time and dose rate that maintains personnel doses

within 10 CFR 20.1201 limits is not exceeded. Therefore, the

proposed changes to the MYTS would 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 location: Wiscasset Public Library, High

Street, PO Box 367, Wiscasset, ME 04578.

Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic

Power Company, PO Box 408, Wiscasset, ME 04578.

NRC Project Director: Seymour H. Weiss.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,

Minnesota.

Date of amendment requests: March 2, 1998.

Description of amendment requests: The proposed amendments would

remove the spent fuel pool special ventilation system operability-based

restriction on crane operations in the spent fuel pool enclosure, while

maintaining that restriction during spent fuel handling operations.

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[s] will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed change does not affect any system that is a

contributor to initiating events for previously evaluated

anticipated operational occurrences and design basis accidents.

Therefore, the proposed change will not increase the probability of

any previously evaluated accident.

The proposed change does not impact the required availability of

the spent fuel pool special ventilation system during spent fuel

handling operations to mitigate the consequences of a fuel handling

accident.

The proposed change does impact the required availability of the

spent fuel pool special ventilation system during heavy load

handling operations. However, this system is not required to

mitigate the consequences of a heavy load dropping onto a spent fuel

assembly. Such a requirement is not applicable at Prairie Island,

because the heavy loads in the spent fuel pool enclosure are either

handled with single-failure-proof cranes, rigging and plant

procedures implementing Prairie Island commitments to NUREG-0612, or

handled with spent fuel pool protective covers in place as described

in the Prairie Island USAR (updated safety analysis report). The use

of a single-failure-proof crane with rigging and procedures that

implement the requirements of NUREG-0612 assures that the potential

for a heavy load

[[Page 27764]]

drop is extremely small and therefore consideration of the effects

of heavy load drops is not required. Spent fuel pool covers prevent

dropped loads* (*The covers do have a limit on the weight load they

are analyzed to withstand.) from falling into the spent fuel pool

and therefore consideration of the effects of heavy load drops is

also not required. These actions taken to reduce the accident

initiator probabilities to insignificant magnitudes negate any

theoretically small increase in the consequence of a postulated

heavy load drop accident resulting from the removal of a requirement

to have one train of the spent fuel pool special ventilation system

operable during crane operations. It is concluded in summary that

the proposed change does not involve a significant increase in the

consequences of any accident previously evaluated.

2. The proposed amendment[s] will not create the possibility of

a new or different kind of accident from any accident previously

analyzed.

The proposed change does impact the required availability of the

spent fuel pool special ventilation system during heavy load

handling operations. Load drop events over spent fuel are well

understood and have been thoroughly evaluated. The proposed change

will not create any new accident scenarios or create the possibility

of a new or different kind of accident from any accident previously

analyzed.

3. The proposed amendment[s] will not involve a significant

reduction in the margin of safety.

The proposed change does not impact the required availability of

the spent fuel pool special ventilation system during spent fuel

handling operations to mitigate the consequences of a fuel handling

accident as described in the USAR. As a result the safety margin

inherent in the 10 CFR part 100 dose limits is not reduced.

The proposed change does impact the required availability of the

spent fuel pool special ventilation system during heavy load

handling operations. However, this system is not required to

mitigate the consequences of a heavy load dropping onto a spent fuel

assembly because the potential for a load drop is extremely small.

Provision of single-failure-proof equipment and compliance with the

other requirements of NUREG-0612 (provide) a defense-in-depth

approach to assure the safe handling of heavy loads which would

otherwise be demonstrated to be safe by the deterministic analysis

of the radiological effects of dropped heavy loads.

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

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW, Washington, DC 20037.

NRC Project Director: Cynthia A. Carpenter.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: November 26, 1997.

Description of amendment request. The amendments to the Units 1 and

2 Technical Specifications Surveillance Requirement Section 4.7.1.3.a

involve lowering the Ultimate Heat Sink (UHS) surveillance requirement

maximum acceptable spray pond average temperature from 88 deg.F to 85

deg.F. This temperature is specified to assure that the post design

basic accident (DBA) loss-of-coolant (LOCA) accident/loss of offsite

power maximum UHS temperature will be maintained less than the UHS

design temperature.

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.

This proposal does not involve an increase in the probability or

consequences of an accident previously evaluated. The proposed

change lowers the UHS temperature surveillance requirement so that

the maximum post DBA UHS temperature is maintained less than that

reported previously.

The UHS provides cooling to equipment and systems required for

the safe shutdown of the plant following an accident with

radiological consequence potential, such as a LOCA. The change in

UHS initial temperature limit to 85 deg.F assures that the peak

temperature will remain less than that reported previously.

Therefore, the components cooled by the UHS will not be impacted and

will be capable of performing their function as designed.

Based upon the analysis presented above, PP&L (Pennsylvania

Power and Light Company) concludes that the proposed action does not

involve an increase in the probability or consequences of an

accident previously evaluated.

2. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

This proposal does not create the probability of a new or

different type of accident from any accident previously evaluated.

The proposed change lowers the UHS surveillance requirement

temperature so that the maximum post DBA UHS temperature is

maintained less than that reported previously. Therefore the

operation of the components cooled by the UHS will not be impacted

and will be capable of performing their design function.

Therefore, the proposed change does not create the possibility

of a new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The change does not involve a reduction in the margin of safety.

The proposed change lowers the UHS surveillance temperature so that

the maximum post DBA UHS temperature is maintained less than that

reported previously. The margin of safety is unaffected since the

maximum post DBA UHS temperature is not affected. Performance of

equipment cooled by the UHS is unaffected.

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: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

and Trowbridge, 2300 N Street NW., Washington, DC 20037.

NRC Project Director: Robert A. Capra.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388,

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: March 16, 1998.

Description of amendment request: The proposed amendment would

change the design basis and Technical Specifications to support the

implementation of Hydrogen Water Chemistry.

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 Design Basis Event requiring functioning of the Main Steam

Line Radiation monitors is defined in the FSAR. FSAR Section

7.2.1.1.4.2.(i) describing Main Steam Line Radiation monitoring

states that for accidents resulting in gross fission

[[Page 27765]]

product release ``the primary variables for trip initiation would be

reactor vessel low level, reactor vessel high pressure, or high

neutron flux''. Because the Main Steam Line Radiation Monitors

[MSLRM] trip function is not used in any accident analyses this

proposed setpoint change does not involve an increase in the

probability or consequences of an accident previously evaluated.

In conformance to SRP 15.4.9, the analysis of the design basis

CRDA assumed release of activity by leakage from an isolated

condenser. As described in the FSAR, the main steam line radiation

monitors will shut down the mechanical vacuum pump if operating and

close its suction valves, thus isolating the condenser in the event

of a Main Steam Line-High Radiation trip. Operation of the

mechanical vacuum pump following burst failures of fuel rods

insufficient to cause a main steam line radiation monitor trip was

evaluated to better understand the potential impacts of raising the

setpoint. Doses calculated under conservative conditions were small

compared to the acceptance criteria for offsite dose of 25% of 10

CFR part 100 limits for offsite dose for the CRDA and 10 CFR 50

limits for control room dose.

Relocation of the Main Condenser Offgas Treatment System

Explosive Gas Monitoring System requirements to the FSAR Section

16.3 (Technical Requirements Manual (TRM)) and procedures involves

the use of an alternate regulatory process for controlling the

instrumentation requirements. The change does not introduce any new

modes of plant operation, make any physical changes, alter any

operational setpoints, or change the surveillance requirements. Any

change in the Main Condenser Offgas Treatment System Explosive Gas

Monitoring System requirements would be evaluated pursuant to the

requirements of 10 CFR 50.59.

The Technical Specifications, the Explosive Gas Mixture

description contained in LCO/Surveillance 3.11.2.6/4.11.2.6 and

associated bases will be moved and retained in TS Section 6.0

``Administrative Controls''. The LCO specific limit and program

details will be relocated to the FSAR Section 16.3 (TRM) and

procedures and any changes controlled by the 10 CFR 50.59 process.

Therefore, this change does not involve an increase in the

probability or consequences of an accident previously evaluated.

These proposed changes to Technical Specifications do not

require physical changes to instrument channels other than the Main

Steam Radiation Monitor setpoint, or to any systems or component

that interfaces with the instrumentation channels, therefore there

is no change in the probability or consequences of any accident

analyzed in the FSAR.

Finally, revising the TS index is an administrative change.

2. Create the possibility of a new or different kind of accident

from any accident previously evaluated.

The proposed Main Steam Line Radiation setpoint change does not

result in any design or physical configuration changes to the

instrumentation channels. Operation incorporating the proposed

change will not impair the instrumentation channels from performing

as provided in the design basis.

Relocation of the Main Condenser Offgas Treatment System

Explosive Gas Monitoring System requirements to the FSAR Section

16.3 (TRM) and procedures involves the use of an alternate

regulatory process for controlling the instrumentation requirements.

Therefore, the above change does not introduce any accident

initiators as it does not involve any new modes of plant operation,

make any physical changes, alter any operational setpoints, or

change the surveillance requirements.

The Technical Specifications, the Explosive Gas Mixture

description contained in LCO/Surveillance 3.11.2.6/4.11.2.6 and

associated bases will be moved and retained in TS Section 6.0

``Administrative Controls''. The LCO specific limit and program

details involves the use of an alternate regulatory process for

controlling the requirements.

Since the proposed changes to the Technical Specifications do

not adversely impact the reliability of the safety required systems,

no new or different kind of accident is created.

3. Involve a significant reduction in a margin of safety.

Raising the trip setpoint does not significantly reduce the

sensitivity of the MSLRM's to alarm and initiate actions in response

to gross fuel failures during power operation or to the design basis

control rod drop accident. The source term assumed for the design

basis CRDA greatly exceeds that required to initiate the main steam

line high radiation trip. Raising the setpoint does not induce a

delay in reaching the setpoint that would result in an increase in

offsite dose from the design basis control rod drop accident. The

delay time from fuel failure to monitor response is determined by

the transport time for steam flow from the reactor vessel to the

monitor location, which is not changed by either hydrogen water

chemistry or by the monitor setpoint. Consequently, raising the trip

setpoint will not result in an incremental increase in activity

release, control room dose or offsite dose. Therefore, there is no

reduction in the margin of safety for the design basis event.

The radiological consequences of small fuel rupture events, that

would produce main steam line radiation levels below the proposed

trip setpoint, are not significant. These postulated events were

evaluated to better understand the potential impacts of raising the

setpoint. The potential offsite doses from such an event, in the

absence of a trip, would be small compared to the limits of 10 CFR

part 50 for control room dose and to the acceptance criteria of 25%

of 10 CFR part 100 limits for offsite dose from the design basis

CRDA.

Relocation of the Main Condenser Offgas Treatment System

Explosive Gas Monitoring System requirements to FSAR Section 16.3

(TRM) involves the use of an alternate regulatory process for

controlling the instrumentation requirements. Any change in the Main

Condenser Offgas Treatment System Explosive Gas Monitoring System

requirements would be evaluated pursuant to the requirements of 10

CFR 50.59. Also, revising the TS index is an administrative change.

The Explosive Gas Mixture description contained in LCO/

Surveillance 3.11.2.6/4.11.2.6 and associated bases will be moved

and retained in TS Section 6.0 ``Administrative Controls''. The LCO

specific limit and program details will be relocated to the FSAR

Section 16.3 (TRM) and procedures and any changes controlled by the

10 CFR 50.59 process.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

and Trowbridge, 2300 N Street NW., Washington, DC 20037.

NRC Project Director: Robert A. Capra.

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: February 16, 1998, as supplemented by

letter dated April 2, 1998.

Description of amendment request: The proposed amendment request

would revise Technical Specification 3/4.4.5, ``Steam Generators,''

and its Bases to allow the implementation of 1-volt voltage-based

repair criteria for the steam generator tube support plate-to-tube

intersections for Unit 2 in accordance with Generic Letter 95-05, and

make related Unit 1 administrative changes for consistency of wording

(the NRC had previously approved a similar 1-volt voltage-based repair

criteria application for Unit 1). In addition, the proposed amendment

would make an administrative change to Bases 3/4.4.6.2,

``Operational Leakage,'' to clarify that the allowable steam generator

leakage specification applies to both Unit 1 and Unit 2.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

[[Page 27766]]

Structural Considerations

Industry testing of model boiler and operating plant tube

specimens for free span tubing at room temperature conditions shows

typical burst pressures in excess of 5000 psi for indications of

ODSCC (outer diameter stress corrosion cracking) with voltage

measurements at or below the current structural limit of 5.45 volts.

One model boiler specimen with a voltage amplitude of 19 volts also

exhibited a burst pressure greater than 5000 psi. Burst testing

performed on one intersection pulled from STP (South Texas Project)

Unit 1 in 1993 with a 0.51 volt indication yielded a measured burst

pressure of 8900 psi at room temperature. Burst testing performed on

another intersection pulled from STP Unit 1 in 1995 with a 0.48 volt

indication yielded a measured burst pressure of 9950 psi at room

temperature.

The next projected end-of-cycle (EOC) voltage compares favorably

with the current structural limit considering the voltage growth

rate for indications at STP. Using the methodology of Generic Letter

95-05, the structural limit is reduced by allowances for uncertainty

and growth to develop a beginning-of-cycle (BOC) repair limit which

should preclude EOC indications from growing in excess of the

structural limit. The non-destructive examination (NDE) uncertainty

to be applied per Generic Letter 95-05 is approximately 20%. The

growth allowance will be 30%/EFPY [effective full power year] or a

STP Unit 2-specific growth rate, to be calculated in accordance with

Generic Letter 95-05, whichever is greater. Where the generator-

specific growth rate exceeds both the Unit 2-specific average growth

rate and 30%/EFPY, that generator-specific growth rate will be used

for that generator. Each succeeding cycle upper voltage repair limit

will also be conservatively established based on Generic Letter 95-

05 methodology. By adding NDE uncertainty allowances and a growth

allowance to the repair limit, the structural limit can be

validated.

The upper voltage repair limit could be applied to bobbin coil

voltages between the lower and upper repair limits to leave such

indications in service independent of RPC [rotating pancake coil]

confirmation. However, RPC-confirmed indications will be

conservatively removed from service consistent with Generic Letter

95-05.

Leakage Considerations

As part of the implementation of voltage-based repair criteria,

the distribution of EOC degradation indications at the TSP (tube

support plate) intersections has been used to calculate the primary-

to-secondary leakage which is bounded by the maximum leakage

required to remain within the applicable dose limits of 10 CFR 100

(10 CFR part 100) and GDC (General Design Criterion) 19. This limit

was calculated using the Technical Specification Reactor Coolant

System (RCS) Iodine-131 transient spiking values consistent with

NUREG-0800. Application of the voltage-based repair criteria

requires the projection of postulated Main Steam Line Break (MSLB)

leakage based on the projected EOC voltage distribution from the

beginning of cycle voltage distribution. Projected EOC voltage

distribution is developed using the most recent EOC eddy current

results and a voltage measurement uncertainty. Draft NUREG-1477 and

Generic Letter 95-05 require that all indications to which voltage-

based repair criteria are applied must be included in the leakage

projection.

The projected MSLB leakage rate calculation methodology

prescribed in Generic Letter 95-05 will be used to calculate the EOC

leakage. A Monte Carlo approach will be used to determine the EOC

leakage, accounting for all of the bobbin coil eddy current test

uncertainties, voltage growth, and an assumed probability of

detection of 0.6. The fitted log-logistic probability of leakage

correlation will be used to establish the MSLB leak rate for each

cycle. This leak rate will be used for comparison with a bounding

allowable leak rate in the faulted loop which would result in

radiological consequences which are within the dose limits of 10 CFR

part 100 for offsite doses and GDC 19 for control room doses. Due to

the relatively low voltage levels of indications at STP to date and

low voltage growth rates, it is expected that the actual calculated

leakage values will be far less than this limit for each successive

cycle.

Other Considerations

Those changes associated with grammatical corrections, deleting

tube diameter information not applicable to South Texas, and

applying the additional reporting requirements to Unit 2, are

administrative and do not involve a change to, or the operation of,

any safety-related system.

Therefore, implementation of voltage-based repair criteria does

not adversely affect steam generator tube integrity and the

radiological consequences will remain below the limits of 10 CFR

part 100 and GDC 19. Operation of the facility in accordance with

the proposed amendment would not result in any increase in the

probability or consequences of an accident previously evaluated.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

Implementation of the proposed steam generator tube voltage-

based repair criteria for ODSCC at the TSP intersections does not

introduce any significant changes to the plant design basis. Use of

the criteria does not provide a mechanism which could result in an

accident outside of the region of the TSP elevations because the

criteria do not apply outside the thickness of the TSPs. It is

therefore expected that for all plant conditions, neither a single

nor multiple tube rupture event would likely occur in a steam

generator where voltage-based repair criteria has been applied.

Specifically, STP has implemented a maximum leakage rate of 150

gpd [gallons-per-day] per steam generator to help preclude the

potential for excessive leakage during all plant conditions. The

draft Reg Guide 1.121 criterion for establishing operational leakage

rate limits governing plant shutdown is based upon leak-before-break

(LBB) considerations to detect a free span crack before potential

tube rupture as a result of faulted plant conditions. The 150 gpd

limit is intended to provide for leakage detection and plant

shutdown in the event of unexpected crack propagation outside the

tube support plate thickness resulting in excessive leakage. Draft

Reg Guide 1.121 acceptance criteria for establishing operating

leakage limits are based on LBB considerations such that plant

shutdown is initiated if permissible degradation is exceeded.

Thus, the 150 gpd limit provides for plant shutdown prior to

reaching critical degradation lengths. Additionally, the leak-

before-break evaluation assumes that the entire crevice area is

uncovered during the secondary side blowdown of a MSLB. Typically,

it is expected for the vast majority of intersections, that only

partial uncovery will occur. Therefore, the proximity of the TSP

will enhance the burst capacity of the tube.

Steam generator tube integrity is continually maintained through

inservice inspection and primary-to-secondary leakage monitoring.

Any tubes falling outside the voltage-based repair criteria limits

are removed from service.

Those changes associated with grammatical corrections, deleting

tube diameter information not applicable to South Texas, and

applying the additional reporting requirements to Unit 2, are

administrative and do not involve a change to, or the operation of,

any safety-related system.

Therefore, operating the facility in accordance with the

proposed amendment would not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The use of the voltage-based bobbin probe for dispositioning

ODSCC degraded tubes within TSP intersections is demonstrated to

maintain steam generator tube integrity in accordance with the

requirements of draft Reg Guide 1.121. Draft Reg Guide 1.121

describes a method acceptable to the NRC staff for meeting GDCs 14,

15, 31, and 32 by reducing the probability or the consequences of

steam generator tube rupture. This is accomplished by determining

the limiting conditions of degradation of steam generator tubing, as

established by inservice inspection, for which tubes with

unacceptable degradation are removed from service. Upon

implementation of the criteria, even under the worst case

conditions, the occurrence of ODSCC at the TSP elevation is not

expected to lead to a steam generator tube rupture event during

normal or faulted plant conditions. The EOC distribution of

indications at the TSP elevations for each successive cycle will be

confirmed to result in acceptable primary-to-secondary leakage

during all plant conditions.

In addressing the combined effects of loss of coolant accident

(LOCA) and safe shutdown earthquake (SSE) on the steam generators,

as required by GDC 2, it has been determined that tube collapse may

occur in the steam generators at some plants. This is not the case

at STP Unit 2 as the TSPs do not become sufficiently deformed as a

result

[[Page 27767]]

of lateral loads at the wedge supports at the periphery of the plate

due to the combined effects of the leak-before-break-limited LOCA

rarefaction wave and SSE loadings to affect tube integrity.

Because the leak-before-break methodology is applicable to the

STP reactor coolant loop piping, the probability of breaks in the

primary loop piping is sufficiently low that they need not be

considered in the structural design of the plant. Implementation

practices using the bobbin probe voltage based tube plugging

criteria bounds Reg Guide 1.83, Rev. 1, considerations by:

(1) Using enhanced eddy current inspection guidelines consistent

with those used by EPRI in developing the correlations. This

provides consistency in voltage normalization.

(2) Performing a 100% bobbin coil inspection for all hot leg

tube support plate intersections and all cold leg intersections down

to the lowest cold leg tube support plate with known ODSCC

indications at each cycle. The determination of the tube support

plate intersections having ODSCC indications shall be based on the

performance of at least a 20% random sampling of tubes inspected

over their full length, and

(3) Incorporating rotating pancake coil inspection for all tubes

with bobbin voltages greater than 1.0 volt. This further establishes

the principal degradation morphology as ODSCC.

Implementation of voltage-based repair criteria at TSP

intersections will decrease the number of tubes which must be

repaired at each subsequent inspection. Since the installation of

tube plugs to remove ODSCC degraded tubes from service reduces the

RCS flow margin, voltage-based repair criteria implementation will

help preserve the margin of flow.

For each cycle the projected EOC primary-to-secondary leak rate

allowed is bounded by a leak rate which limits the radiological

consequences of a EOC MSLB to within the dose limits of 10CFR100 for

offsite doses and 10CFR50, Appendix A, General Design Criteria (GDC)

19 for control room doses. Therefore, this change does not involve a

significant reduction in the margin to safety.

The assessment of radiological consequences of an assumed steam

line break applicable to STP Unit 1 was provided in Attachment 2 to

ST-HL-AE-5359 on May 2, 1996. The submittal was made in response to

questions from the Emergency Preparedness and Radiation Protection

Branch and is applicable to Unit 2 as well. The staff concluded that

the thyroid doses for the Exclusion Area Boundary (EAB), Low

Population Zone (LPZ), and control room are within the acceptance

criteria.

Those changes associated with grammatical corrections, deleting

tube diameter information not applicable to South Texas, and

applying the additional reporting requirements to Unit 2, are

administrative and do not involve a change to, or the operation of,

any safety-related system.

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, NW., Washington, DC 20036-5869.

NRC Project Director: John N. Hannon.

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.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North

Carolina

Date of amendment request: April 24, 1998.

Brief description of amendment: The proposed amendment would revise

Technical Specification 3/4.3.2, ``Engineered Safety Features Actuation

System Instrumentation'' to allow a 2-hour surveillance interval to

facilitate testing of the 6.9 kV Emergency Bus Undervoltage relays.

Date of publication of individual notice in the Federal Register:

May 4, 1998 (63 FR 24574).

Expiration date of individual notice: May 18, 1998 for comments;

June 3, 1998 for hearings.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Northeast Nuclear Energy Company, Docket No. 50-423, Millstone Nuclear

Power Station, Unit No. 3, New London, County, Connecticut

Date of amendment request: April 7, 1998.

Description of amendment request: The proposed amendment would

replace the pressurizer maximizer water inventory requirement with a

pressurizer maximizer indicated level requirement.

Date of publication of individual notice in Federal Register: April

23, 1998 (63 FR 20219)

Expiration date of individual notice: May 26, 1998.

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, Docket No. 50-423, Millstone Nuclear

Power Station, Unit 3, New London, County, Connecticut

Date of amendment request: April 14, 1998.

Description of amendment request: The proposed amendment addresses

an earlier identified condition relating to the plant operators'

ability to meet the operator response time of 10 minutes assumed in

Chapter 15 of the Final Safety Analysis Report for termination of an

Inadvertent Safety Injection event.

Date of publication of individual notice in Federal Register: April

20, 1998 (63 FR 19532).

Expiration date of individual notice: May 20, 1998.

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.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: August 1, 1996, as supplemented on March

2, 1998.

Brief description of amendment request: The proposed amendments

would revise the Technical Specifications as follows: (1.n.) Change the

surveillance requirement frequency for verification that the average

planar

[[Page 27768]]

heat generation rate, minimum critical power ratio, linear heat

generation rate, and average power range monitor gain and setpoint are

within specified limits. Specifically, the frequency would be changed

from within 12 hours after completion of a thermal power increase of at

least 15 percent of rated thermal power (RTP) to once within 24 hours

after greater than or equal to 25 percent RTP, 24 hours thereafter, and

prior to exceeding 50 percent RTP; (2.o.) Change the surveillance

requirement for the verification of the average power range monitor

flow biased simulated thermal power-high time constant from 6 seconds

plus or minus 1 second to less than 7 seconds. The lower limit of 5

seconds will be relocated to plant procedures since it is not a

condition for operability of this reactor protection system function;

(3.p.) Change the frequency of surveillance requirement for rod worth

minimizer channel functional test; and (4.q.) Relocate the main steam

line radiation monitor reactor protection system and isolation trips

from the Technical Specifications to the plant-controlled Technical

Requirements Manual.

Date of publication of individual notice in Federal Register: 1.n.

April 27, 1998 (63 FR 20664); 2.o. April 27, 1998 (63 FR 20669); 3.p.

April 27, 1998 (63 FR 20665); 4.q. April 27, 1998 (63 FR 20667).

Expiration date of individual notices: May 27, 1998 (all 4

notices).

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

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: February 27, 1998.

Brief Description of amendment: The amendment revised the Technical

Specifications by revising the pressure-temperature and overpressure

limits.

Date of publication of individual notice in Federal Register: March

9, 1998 (63 FR 11456).

Expiration date of individual notice: April 8, 1998.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,

Unit 1, Rhea County, Tennessee

Date of application for amendment: April 29, 1998.

Brief description of amendments: To amend the Watts Bar Nuclear

Plant, Unit 1, Technical Specifications (TS) for the Hydrogen

Mitigation System igniters. The amendment revises the TS limiting

condition for operation, LCO 3.6.8, to provide temporary requirements

for hydrogen ignitors to address the two Train A ignitors which are

currently out of service.

Date of publication of individual notice in the Federal Register:

May 7, 1998 (63 FR 25243).

Expiration date of individual notice: June 8, 1998.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402.

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 Ch. 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.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,

Maryland

Date of application for amendments: December 4, 1996, as

supplemented March 27, June 9, June 18, July 21, August 14, August 19,

September 10, October 6, October 20, October 23, November 5, 1997, and

January 12, January 28 and March 16, 1998.

Brief description of amendments: The amendments include the

following:

1. The amendments added a new surveillance requirement (SR) 3.4.9.2

to the Improved Technical Specifications (ITS) which requires

verification that the capacity of each required bank of pressurizer

heaters is equal to or greater than 150 kW every 24 months.

2. The amendments changed the current TS applicability for the

pressurizer safety valves for Mode 3 to specify that two safety valves

shall be operable with all reactor coolant system (RCS) cold leg

temperature 365 deg.F for Unit 1 and >301 deg.F for Unit

2. This is a less restrictive change.

3. As part of the conversion to the ITS, the amemdment changed a

requirement that the power-operated relief valves be demonstrated

operable by performing a channel functional test once per 31 days to

once per 92 days.

4. The ITS LCO 3.4.1.3 eliminated the limit of 1 gpm total primary-

to-secondary leakage through all steam generators and thus will only

require a limit of 100 gallons per day through any one steam generator.

This is an administrative change.

5. The amendment retains the requirement of SR 4.5.2.f.2 and

specifies a frequency of 24 months. The amendment also adds a new SR

3.5.2.7 which requires verification that each LPSI pump stops on an

actual or simulated actuation signal.

6. The amendment regarding the control room emergency ventilation

system (CREVS) changes the surveillance interval from 18 months to 24

months (each refueling cycle) for SR 4.7.6.1.e.2 requires that each

train of CREVS is demonstrated operable at least once every 18 months

by verifying that on a control room high radiation test signal, the

system automatically switches into a recirculation mode of operation

with flow through the HEPA filter and charcoal adsorber banks and that

both of the isolation valves in each duct and common exhaust duct, and

isolation valve in the toilet exhaust area duct, close. The above

change is less restrictive.

[[Page 27769]]

7. The amendment changes the surveillance interval regarding the

control room emergency temperature system (CRETS) from 62 days on a

staggered basis (one train every 31 days) to 24 months (each refueling

interval) for SR 4.7.6.1.a.

8. The amendment changes the surveillance interval regarding the

spent fuel pool exhaust ventilation system (SFPEVS) from 18 months to

24 months (each refueling interval) for SR 4.9.12.d. This is a less

restrictive change.

9. The amendment changes the surveillance interval regarding the

penetration room exhaust ventilation system (PREVS) from 18 months to

24 months (each refueling interval) for SR 4.6.6.1.d.2.

Date of issuance: May 4, 1998.

Effective date: As of the date of issuance to be implemented by

August 31, 1998.

Amendment Nos.: 227 and 201.

Facility Operating License Nos. DPR-53 and DPR-69: Amendments

revised the Technical Specifications in its entirety.

Date of initial notice in Federal Register: March 6, 1998 (63 FR

11312) The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated May 4, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of application for amendments: October 29, 1997, as

supplemented January 28 and April 20, 1998.

Brief Description of amendments: The amendments update the

Technical Specification description of Control Rod Assemblies to allow

for boron carbide or hafnium absorber materials, as approved by the NRC

staff.

Date of issuance: April 27, 1998.

Effective date: April 27, 1998.

Amendment Nos.: 193 and 224.

Facility Operating License Nos. DPR-71 and DPR-62: Amendments

change the Technical Specifications.

Date of initial notice in Federal Register: December 17, 1997 (62

FR 66137) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 27, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Date of application for amendments: November 7, 1997, as

supplemented on March 24, 1998, and April 9, 1998.

Brief description of amendments: The amendments defer the next

scheduled Type A containment integrated leak rate test for Byron, Unit

2, until the next refueling outage in 1999.

Date of issuance: May 8, 1998.

Effective date: Immediately, to be implemented within 30 days.

Amendment Nos.: 102 and 102.

Facility Operating License Nos. NPF-37 and NPF-66: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: April 7, 1998 (63 FR

17036) The April 9, 1998, supplement provided clarifying information

which did not change the staff's initial proposed no significant

hazards consideration.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 8, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Byron Public Library District,

109 N. Franklin, P.O. Box 434, Byron, Illinois 61010.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle County, Illinois

Date of application for amendments: September 26, 1997, as

supplemented on April 7, 1998.

Brief description of amendments: The amendments revise Technical

Specification 3.6.1.8 to prohibit the simultaneous opening of the

drywell and suppression chamber purge system isolation valves and

revise the surveillance requirements of TS 3/4.6.5.3, ``Standby Gas

Treatment System'' to upgrade the filter testing methods to more

current industry standards. This amendment approves only a portion of

the request dated September 26, 1997. The remainder of the request will

be addressed in separate correspondence.

Date of issuance: April 27, 1998.

Effective date: Immediately, to be implemented prior to startup of

LaSalle, Unit 1, from the current outage and prior to restart of

LaSalle, Unit 2, from the current outage.

Amendment Nos.: 125 and 110.

Facility Operating License Nos. NPF-11 and NPF-18: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: November 19, 1997 (62

FR 61840) The April 7, 1998, 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

April 27, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

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

County, Michigan

Date of application for amendment: December 11, 1995, as

supplemented January 18, September 3, October 2, October 18, and

October 25, 1996, and March 28, 1997.

Brief description of amendment: The amendment revises

administrative controls technical specifications (TS) and related

surveillance requirements. Amendment 174, issued on October 31, 1996,

provided a partial response to the licensee's request. This amendment

completes action on the request.

NRC has also granted the request of Consumers Energy to withdraw a

portion of its December 11, 1996, application. The proposed change

would have deleted the requirements of current TS 4.5.4, ``Surveillance

for Prestressing System,'' TS 4.5.5, ``End Anchorage Concrete

Surveillance,'' and TS 4.5.8, ``Dome Delamination Surveillance,'' and

replaced the requirements with proposed TS 6.5.5, ``Containment

Structural Integrity Surveillance Program.'' However, by letter dated

March 28, 1997, the licensee withdrew the proposed change. In addition,

the staff has denied a portion of the amendment request regarding

limitations on the dose rates resulting from radioactive material

released in gaseous effluents to areas beyond the site boundary. A

separate Notice of Partial Denial of Amendment to Facility Operating

License and Opportunity for Hearing has been published in the Federal

Register. For further details with respect to these actions, see the

application for amendment dated December 11, 1996, as supplemented

above, the licensee's letter dated March 28, 1997, which withdrew this

portion of the application for license amendment, and the staff's

Safety Evaluation enclosed with the

[[Page 27770]]

amendment. The above documents 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 room

listed below.

Date of issuance: May 7, 1998.

Effective date: May 7, 1998, to be implemented within 60 days from

date of issuance.

Amendment No.: 181.

Facility Operating License No. DPR-20: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 20, 1996 (61

FR 49493) The October 2, October 18, and October 25, 1996, and March

28, 1997, letters provided clarifying information and updated TS pages

that were within the scope of the original Federal Register notice and

did not change the staff's initial proposed no significant hazards

considerations determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 7, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Van Wylen Library, Hope

College, Holland Michigan 49423.

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

County, Michigan

Date of application for amendment: December 27, 1995, as

supplemented September 4, October 18, and November 26, 1996, June 27

and November 21, 1997, and January 29, and April 10, 1998.

Brief description of amendment: The amendment revises specification

requirements and associated bases regarding the electrical power

systems to closely emulate the Standard Technical Specifications for

Combustion Engineer Plants, NUREG-1432, Revision 1.

Date of issuance: April 29, 1998.

Effective date: The license amendment is effective as of the date

of issuance with full implementation within 60 days after Cold Shutdown

following completion of the 1998 refueling outage, but no later than

October 2, 1998.

Amendment No.: 180.

Facility Operating License No. DPR-20: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 9, 1997 (62 FR

17229) The June 27 and November 21, 1997, and January 29 and April 10,

1998, letters provided clarifying information that was within the scope

of the original Federal Register notice and did not change the staff's

initial proposed no significant hazards considerations determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 29, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Entergy Mississippi, Inc.,

Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of application for amendment: September 18, 1997, as

supplemented by letter dated February 24, 1998.

Brief description of amendment: The amendment decreases the safety

limit for the minimum critical power ratio (MCPR) from 1.12 to 1.11 for

two recirculation loop operation and from 1.14 to 1.12 for single

recirculation loop operation in Technical Specification (TS) 2.1.1.2.

Because the proposed amendment is for Cycle 10 operation, the amendment

would also revise the footnotes to TSs 2.1.1.2 and 5.6.5 to state that

the MCPR values and the items 19 and 20, two topical reports being

added to the core operating limits report in TS 5.6.5, are ``applicable

only for Cycle 10 operation.'' Cycle 10 operation begins at the plant

restart from the current refueling outage No. 9.

Date of issuance: May 8, 1998.

Effective date: May 8, 1998.

Amendment No: 136.

Facility Operating License No. NPF-29: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: October 22, 1998 (62 FR

54872) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated May 8, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Judge George W. Armstrong

Library, 220 S. Commerce Street, Natchez, MS 39120.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: October 4, 1997.

Brief description of amendment: To revise the Final Safety Analysis

Report (FSAR) and the Improved Technical Specification (TS) Bases to

reflect the modified reactor building fan control logic for fan AHF-1C.

Date of issuance: April 29, 1998.

Effective date: April 29, 1998.

Amendment No.: 166.

Facility Operating License No. DPR-72: Amendment revised the

updated FSAR and TS Bases.

Date of initial notice in Federal Register: November 13, 1997 (62

FR 60921) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 29, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629.

Florida Power and Light Company, et al., Docket No. 50-335, St. Lucie

Plant, Unit No. 1, St. Lucie County, Florida

Date of application for amendment: July 22, 1997.

Brief description of amendment: The amendment will incorporate a

recent evaluation of a postulated inadvertent opening of a main steam

safety valve into the current licensing basis for St. Lucie Unit 1.

Date of Issuance: April 30, 1998.

Effective Date: April 30, 1998.

Amendment No.: 154.

Facility Operating License No. NPF-16: Amendment revised the

Updated Final Safety Evaluation Report.

Date of initial notice in Federal Register: August 27, 1997 (62 FR

45457) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 30, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Community College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,

Minnesota

Date of application for amendments: March 6, 1998, as supplemented

March 30, March 31, and April 13, 1998.

Brief description of amendments: The amendments update the

Technical Specification heatup and cooldown rate curves and extend

their reactor vessel fluence limit from the current 20 effective full

power years (EFPYs) to a new value of 35 EFPYs, incorporate into

Technical Specifications the use of a Pressure and Temperature Limits

Report, and change the power-operated relief valves temperature

requirement for operability.

Date of issuance: May 4, 1998.

Effective date: May 4, 1998, with full implementation within 30

days.

[[Page 27771]]

Amendment Nos.: 135, 127.

Facility Operating License Nos. DPR-42 and DPR-60. Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: March 27, 1998 (63 FR

14972) The March 30, March 31, and April 13, 1998, letters provided

clarifying information and updated Technical Specification pages within

the scope of the original Federal Register notice and did not change

the staff's initial proposed no significant hazards considerations

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 4, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Philadelphia Electric Company, Docket No. 50-352, Limerick Generating

Station, Unit 1, Montgomery County, Pennsylvania.

Date of application for amendment: February 9, 1998, as

supplemented April 8 and 24, 1998.

Brief description of amendment: The amendment revises the minimal

critical power ratio safety limits for operation Cycle 8.

Date of issuance: May 4, 1998.

Effective date: As of date of issuance, and shall be implemented

within 30 days.

Amendment No.: 127.

Facility Operating License No. NPF-39: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 25, 1998 (63

FR 9613) The April 8 and 24, 1998, letters provided clarifying

information that did not change the initial proposed no significant

hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 4, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-

364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,

Alabama.

Date of amendments request: February 14, 1997, as supplemented by

letters dated June 20, August 5, September 22, November 19, December 9,

December 17, and December 31, 1997, January 23, February 12, February

26, March 3, March 6, March 16, April 3, April 13, and two letters on

April 17, 1998.

Brief Description of amendments: The amendments change the maximum

reactor core power level for facility operation from 2652 megawatts-

thermal (MWt) to 2775 MWt for the Joseph M. Farley Nuclear Plant, Units

1 and 2. The amendments also approve changes to the Technical

Specifications to implement uprated power operation.

Date of issuance: April 29, 1998.

Effective date: As of the date of issuance to be implemented prior

to entering Mode 4 for Cycle 16 (fall 1998) for Unit 1 and prior to

entering Mode 4 for Cycle 13 (spring 1998) for Unit 2.

Amendment Nos.: Unit 1--137; Unit 2--129.

Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise

the Technical Specifications, Operating Licenses, and adds a new

Appendix C to the Operating Licenses.

Date of initial notice in Federal Register: October 8, 1997 (62 FR

52588) The November 19, December 9, December 17, and December 31, 1997,

January 23, February 12, February 26, March 3, March 6, March 16, April

3, April 13, and two letters on April 17, 1998, provided additional and

clarifying information that did not change the scope of the February

14, 1997, 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 April 29, 1998, and an Environmental

Statement was prepared and dated April 17, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Houston-Love Memorial Library,

212 W. Berdache Street, Post Office Box 1369, Dothan, Alabama.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2, Somervell County, Texas.

Date of amendment request: April 9, 1998 (TXX-98107).

Brief description of amendments: The proposed amendment would allow

on a one time basis, the verification of the proper operation of the

Unit 2 load shed seal-in contacts and the diesel generator trip bypass

contacts at power and crediting performance of Surveillance

Requirements (SR) 4.8.1.1.2f.4) and 4.8.1.1.2f.6), at power as opposed

to ``during shutdown'' as currently required by those SR. The proposed

amendment would also allow on a one time basis the verification of the

proper operation of the Unit 2 lockout relays and contacts to be

deferred until the startup from the Unit 2 fourth refueling outage

(2RFO4) or earlier outage to at least MODE 3.

Date of issuance: May 8, 1998.

Effective date: May 8, 1998.

Amendment Nos.: Unit 1--Amendment No. 59; Unit 2--Amendment No. 45.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: April 20, 1998, (63 FR

19534). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 8, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, TX 76019.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri.

Date of application for amendment: August 8, 1997, as supplemented

by letter dated November 10, 1997.

Brief description of amendment: The amendment revises the feedwater

isolation engineered safety feature actuation system (ESFAS) functions

in Technical Specification Tables 3.3-3, 3.3-4, and 4.3-2.

Date of issuance: April 23, 1998.

Effective date: April 23, 1998, to be implemented within 30 days

from the date of issuance.

Amendment No.: 126.

Facility Operating License No. NPF-30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 17, 1997 (62

FR 66144) The November 10, 1997, supplemental letter provided

additional clarifying information that did not change the staff's

original no significant hazards consideration determination. The

Commission's related evaluation of the amendment is contained in a

Safety Evaluation dated April 23, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Missouri--

Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek

Generating Station, Coffey County, Kansas

Date of amendment request: February 4, 1998.

[[Page 27772]]

Brief description of amendment: The amendment would revise

Technical Specification 3.2.4, ``Quadrant Power Tilt Ratio,'' (QPTR)

and its associated Bases to reflect (1) a change in the action for

determining QPTR when QPTR is above 1.02, (2) a change in the

completion time for resetting the power range neutron flux-high trip

setpoints after QPTR is determined to be above 1.02, and (3) deletion

of actions requiring QPTR to be restored within 24 hours, QPTR to be

verified during a return to power operation, resetting the power range

neutron flux-high trip setpoint to less than 55 percent following a

power reduction to 50 percent reactor thermal power or below, and

actions for QPTR in excess of 1.09.

Date of issuance: April 27, 1998.

Effective date: April 27, 1998, to be implemented within 60 days

from the date of issuance.

Amendment No.: 116.

Facility Operating License No. NPF-42: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 25, 1998 (63 FR

14489) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 27, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621.

Dated at Rockville, MD., this 13th day of May 1998.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects--III/IV, Office of

Nuclear Reactor Regulation.

[FR Doc. 98-13223 Filed 5-19-98; 8:45 am]

BILLING CODE 7590-01-P