Federal Register / Vol. 63, No. 97 /Wednesday, May 20, 1998 / Notices 27757

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.
27758 Federal Register / Vol. 63, No. 97 /Wednesday, May 20, 1998 / Notices

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.
Federal Register / Vol. 63, No. 97 /Wednesday, May 20, 1998 / Notices 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
27760 Federal Register / Vol. 63, No. 97 /Wednesday, May 20, 1998 / Notices

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.
Federal Register / Vol. 63, No. 97 /Wednesday, May 20, 1998 / Notices 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
27762 Federal Register / Vol. 63, No. 97 /Wednesday, May 20, 1998 / Notices

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
Federal Register / Vol. 63, No. 97 /Wednesday, May 20, 1998 / Notices 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-failureproof crane with rigging and procedures that

implement the requirements of NUREG–0612

assures that the potential for a heavy load
27764 Federal Register / Vol. 63, No. 97 /Wednesday, May 20, 1998 / Notices

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 singlefailure-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 °F to 85 °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 °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
Federal Register / Vol. 63, No. 97 /Wednesday, May 20, 1998 / Notices 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-totube 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.
27766 Federal Register / Vol. 63, No. 97 /Wednesday, May 20, 1998 / Notices

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 generatorspecific 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 voltagebased 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 voltagebased 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 [gallonsper-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 leakbefore-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
Federal Register / Vol. 63, No. 97 /Wednesday, May 20, 1998 / Notices 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 primaryto-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
27768 Federal Register / Vol. 63, No. 97 /Wednesday, May 20, 1998 / Notices

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 pressuretemperature 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 °F for

Unit 1 and >301 °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-tosecondary 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.
Federal Register / Vol. 63, No. 97 /Wednesday, May 20, 1998 / Notices 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
27770 Federal Register / Vol. 63, No. 97 /Wednesday, May 20, 1998 / Notices

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.
Federal Register / Vol. 63, No. 97 /Wednesday, May 20, 1998 / Notices 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.
27772 Federal Register / Vol. 63, No. 97 /Wednesday, May 20, 1998 / Notices

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

OFFICE OF MANAGEMENT AND

BUDGET

Cumulative Report on Rescissions and

Deferrals

May 1, 1998.

This report is submitted in fulfillment

of the requirement of Section 1014(e) of

the Congressional Budget and

Impoundment Control Act of 1974

(Public Law 93–344). Section 1014(e)

requires a monthly report listing all

budget authority for the current fiscal

year for which, as of the first day of the

month, a special message had been

transmitted to Congress.

This report gives the status, as of May

1, 1998, of 24 rescission proposals and

eight deferrals contained in two special

messages for FY 1998. These messages

were transmitted to Congress on

February 3 and February 20, 1998.

Rescissions (Attachments A and C)

As of May 1, 1998, 24 rescission

proposals totaling $20 million had been

transmitted to the Congress. Congress

approved 21 of the Administration’s

rescission proposals in P.L. 105–174. A

total of $17.3 million of the rescissions

proposed by the President was

rescinded by that measure. Attachment

C shows the status of the FY 1998

rescission proposals.

Deferrals (Attachments B and D)

As of May 1, 1998, $3,293 million in

budget authority was being deferred

from obligation. Attachment D shows

the status of each deferral reported

during FY 1998.

Information From Special Messages

The special messages containing

information on the rescission proposals

and deferrals that are covered by this

cumulative report are printed in the

editions of the Federal Register cited

below:

63 FR 7004, Wednesday, February 11,

1998

63 FR 10076, Friday, February 27, 1998

Franklin D. Raines,

Director.

Attachments

ATTACHMENT A.—STATUS OF FY 1998 RESCISSIONS

[In millions of dollars]

Budgetary

resources

Rescissions proposed by the President .................................................................................................................................................. 20.1

Rejected by the Congress .......................................................................................................................................................................

Amounts rescinded by P.L. 105–174, the FY 1998 Supplemental Appropriations and Rescissions Act ............................................... ¥17.3

Currently before the Congress ................................................................................................................................................................ 2.8

ATTACHMENT B.—STATUS OF FY 1998 DEFERRALS

[In millions of dollars]

Budgetary

resources

Deferrals proposed by the President ....................................................................................................................................................... 4,833.0

Routine Executive releases through May 1, 1998 (OMB/Agency releases of $1,540.1 million, partially offset by cumulative positive

adjustment of $0.3 million) ................................................................................................................................................................... ¥1,539.8

Overturned by the Congress ................................................................................................................................................................... ....................

Currently before the Congress ................................................................................................................................................................ 3,293.2