Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices 11125

Dated at Rockville, Maryland, this 22nd

day of February 1995.

For the Nuclear Regulatory Commission.

Ronald W. Hernan,

Acting Director, Project Directorate I–4,

Division of Reactor Projects—I/II, Office of

Nuclear Reactor Regulation.

[FR Doc. 95–4975 Filed 2–28–95; 8:45 am]

BILLING CODE 7590–01–M

[Docket No. 50–413]

Duke Power Company, et al., Catawba

Nuclear Station, Unit No. 1;

Environmental Assessment and

Finding of No Significant Impact

The U.S. Nuclear Regulatory

Commission (the Commission) is

considering issuance of an exemption

from certain requirements of 10 CFR

Part 50, Appendix J, Paragraph

III.D.1.(a), Type A Tests, to the Duke

Power Company, et al. (the licensee), for

operation of the Catawba Nuclear

Station, Unit No. 1, located in York

County, South Carolina, in accordance

with Facility Operating License No.

NFP–35.

Environmental Assessment

Identification of the Proposed Action

This Environmental Assessment has

been prepared to address potential

environmental issues related to the

licensee’s application of October 18,

1994, as supplemented on February 7,

1995. The proposed action would

exempt the licensee from the

requirements of 10 CFR Part 50,

Appendix J, Paragraph III.D.1.(a), to the

extent that a one-time schedular

extension would permit rescheduling

the third containment integrated leak

rate test (ILRT) in the first 10-year

service period from the end-of-Cycle 8

outage until the end-of-Cycle 9 outage.

The requested exemption would also

allow the decoupling of this third test

from the endpoint of the first 10-year

inservice inspection.

The Need for the Proposed Action

The current containment integrated

leakage rate (ILRT) requirements for

Catawba Units 1 and 2, pursuant to

Appendix J, are that, after the

preoperational leak rate test, a set of

three Type A tests must be performed at

approximately equal intervals during

each 10-year period. Also, the third test

of each set must be conducted when the

plant is shut down for the 10-year plant

inservice inspection. This is reflected in

the Catawba Technical Specifications

(TS) as a testing interval of once each 40

months plus or minus 10 months, for a

frequency of three times in a 120-month

period. To date, for Catawba Unit 1, the

preoperational and the first two periodic

ILRTs have been conducted. The most

recent ILRT was conducted in March

1991, approximately 47 months ago.

Thus, in accordance with Appendix J

and the current TS, and ILRT would

have to be conducted during the

refueling outage beginning in February

1995 (the end-of-cycle (EOC) 8 outage).

The licensee has requested an

exemption from Appendix J and a

corresponding change to the TS that

would allow a one-time change to the

interval for the Unit 1 ILRT from 40 plus

or minus 10 months to 60 plus or minus

10 months (once each 5 years). This

would allow the EOC–8 ILRT to be

rescheduled for EOC–9. Therefore, the

need for the licensee’s proposed action

is to allow a longer interval between the

Catawba Unit 1 second and third

periodic Type A ILRTs which will result

in a cost savings to the licensee.

Environmental Impacts of the Proposed

Action

The proposed one-time exemption

would not increase the probability or

consequences of accidents previously

analyzed and the proposed one-time

exemption would not affect facility

radiation levels or facility radiological

effluents. The licensee has analyzed the

results of previous Type A tests

performed at the Catawba Nuclear

Station, Unit No. 1. The licensee has

provided an acceptable basis for

concluding that the proposed one-time

extension of the Type A test interval

would maintain the containment

leakage rates within acceptable limits.

Accordingly, the Commission has

concluded that the one-time extension

does not result in a significant increase

in the amounts of any effluents that may

be released nor does it result in a

significant increase in individual or

cumulative occupational radiation

exposure. Therefore, there are no

significant radiological environmental

impacts associated with the proposed

exemption.

With regard to potential

nonradiological impacts, the proposed

exemption only involves Type A testing

on the containment. It does not affect

nonradiological plant effluents and has

no other environmental impact.

Accordingly, the Commission concludes

that there are no significant

nonradiological environmental impacts

associated with the proposed

exemption.

Alternatives to the Proposed Action

Since the Commission has concluded

there is no measurable environmental

impact associated with the proposed

exemption, any alternatives with equal

or greater environmental impact need

not be evaluated. The principal

alternative to this action would be to

deny the request for exemption. Such

action would not reduce the

environmental impacts of plant

operations.

Alternative Use of Resources

This action does not involve the use

of resources not previously considered

in the ‘‘Final Environmental Statement

Related to the Operation of Catawba

Nuclear Station Unit No. 1,’’ dated

January 1983.

Agencies and Persons Consulted

In accordance with its stated policy,

the NRC staff consulted with the South

Carolina State official regarding the

environmental impact of the proposed

action. The State official had no

comments.

Finding of No Significant Impact

Based upon the environmental

assessment, the Commission concludes

that the proposed action will not have

a significant effect on the quality of the

human environment. Accordingly, the

Commission has determined not to

prepare an environmental impact

statement for the proposed exemption.

For further details with respect to this

action, see the licensee’s letter dated

October 18, 1994, as supplemented

February 7, 1995, which 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 located at the

York County Library, 138 East Black

Street, Rock Hill, South Carolina.

Dated at Rockville, Maryland, this 23rd day

of February 1995.

For the Nuclear Regulatory Commission.

Herbert N. Berkow,

Director, Project Directorate II–3, Division of

Reactor Projects—I/II, Office of Nuclear

Reactor Regulation.

[FR Doc. 95–4976 Filed 2–28–95; 8:45 am]

BILLING CODE 7590–01–M

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

I. Background

Pursuant to Pubic Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission on NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the
11126 Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices

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 February 3,

1995, through February 16, 1995. The

last biweekly notice was published on

February 15, 1995 (60 FR 8739).

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 Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, and should cite

the publication date and page number of

this Federal Register notice. Written

comments may also be delivered to

Room 6D22, Two White Flint North,

11545 Rockville Pike, Rockville,

Maryland from 7:30 a.m. to 4:15 p.m.

Federal workdays. Copies of written

comments received may be examined at

the NRC Public Document Room, the

Gelman Building, 2120 L Street, NW.,

Washington, DC. The filing of requests

for a hearing and petitions for leave to

intervene is discussed below.

By March 31, 1995, 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 shale be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no
Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices 11127

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, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

Union at 1–(800) 248–5100 (in Missouri

1–(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, 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

section, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Unit Nos. 1, 2, and

3, Maricopa County, Arizona

Date of amendment requests:

December 7, 1994.

Description of amendment requests:

The proposed amendment would revise

the capacity of the ultimate heat sink

(UHS) as described in the bases of

Technical Specification 3/4.7.5,

‘‘Ultimate Heat Sink,’’ from providing a

27-day cooling water supply to

providing a 26-day cooling water

supply. In addition, the reference to

Regulatory Guide 1.27 in the bases of

this TS would also be revised to

reference the January 1976 revision

rather than the March 1974 revision.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensees have provided their analysis

about the issue of no significant hazards

consideration, which is presented

below:

Standard 1—Does the proposed change

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

The Essential spray pond system and the

UHS do not initiate any accidents in

Chapters 6 or 15 of the UFSAR [Updated

Final Safety Analysis Report]. The

justification and basis for the time that the

UHS is available is not changed and

continues to be consistent with the guidance

in Regulatory Guide 1.27. The existing

Technical Specification requirements and

those components to which they apply are

not altered by this Technical Specification

amendment. Therefore, the change to the

bases for Technical Specification 3/4.7.5 does

not increase the probability of occurrence or

the consequences of any previously

evaluated accident.

Standard 2—Does the proposed change

create the possibility of a new or different

kind of accident from any accident

previously evaluated?

The requirements for Technical

Specification 3/4.7.5 are not changed. This

amendment has no impact on plant

maintenance, testing, shutdown equipment,

or component qualification. Therefore, the

possibility of a new or different kind of

accident is not created by this amendment.

Standard 3—Does the proposed change

involve a significant reduction in a margin of

safety?

The change to the bases for Technical

Specification 3/4.7.5 does not significantly

alter existing Technical Specification

requirements or those coponments to which

they apply. The justification and basis for the

time that the UHS is available without

makeup is not changed and continues to be

consistent with the guidance in Regulatory

Guide 1.27. Regulatory Guide 1.27 states that

‘‘A capacity less than 30 days may be

acceptable if it can be demonstrated that

replenishment can be effected to ensure that

continuous capability of the sink to perform

its safety functions, taking into account the

availability of replenishment equipment and

limitations that may be imposed on ‘‘freedom

of movement’’ following an accident.’’ This

change does not effect the continuous

capability of the UHS to perform its safety

function of providing decay heat removal

capability following an accident. The change

updates the design basis of the UHS using

more realistic conditions based on plant

experience. Therefore, the change in the

capacity of the UHS without makeup from 27

days to 26 days will not involve a significant

reduction in margin of safety for the ultimate

heat sink.

The NRC staff has reviewed the

licensees’ analysis and, based on that

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards

consideration.

Local Public Document Room

location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona

85004.

Attorney for licensees: Nancy C.

Loftin, Esq., Corporation Secretary and

Counsel, Arizona Public Service

Company, P.O. Box 53999, Mail Station

9068, Phoenix, Arizona 85072–3999.

NRC Project Director: Theodore R.

Quay.

Carolina Power & Light Company,

Docket No. 50–261, H. B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of amendment request: June 18,

1992, as supplemented December 8,

1992, and revised February 3, 1995.

Description of amendment request:

The proposed Technical Specification

(TS) amendment adds limiting

conditions of operation and surveillance

requirements for the pressurizer poweroperated relief valves (PORVs) and their

associated block valves whenever

average temperature (Tavg) is above 350

degrees F or the reactor is critical.

Specifications have also been added for

low-temperature overpressure

protection whenever Tavg is less than

350 degrees F and the reactor coolant

system is not vented to the containment.

The February 3, 1995, revision made

editorial changes to previous TS pages

and made changes to conform with an

additional provision of the guidance for

surveillance testing of the block valves

associated with the pressurizer PORVs.

In addition, the licensee has requested

an editorial change to TS page 3.1.–11

to revise the references to two figures

that have been superseded.

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 requested revision does not involve

a significant increase in the probability or

consequences of an accident previously
11128 Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices

evaluated. The proposed revision to our

previous Technical Specification (TS) change

request dated June 18, 1992, would help

assure the availability of the block valves for

accident mitigation. The availability of the

block valves for accident mitigation has been

found to outweigh any negative safety

consequences associated with full cycle

testing of a block valve isolating a pressurizer

power-operated relief valves (PORV) with

‘‘excessive’’ seat leakage. There would be no

significant increase in the probability or

consequences of an accident previously

evaluated since this event is fully bounded

by the failing open of a single pressurizer

code safety relief valve event which is

analyzed in Chapter 15 of the Updated Final

Safety Analysis Report. Accordingly, the

requested revision will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The requested revision to our previous

TS change request does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. Periodic testing of the block valves

in accordance with the requested revision is

only intended to assure the functioning and

capability of the block valves. The requested

revision will only clarify the conditions

when block valve surveillance testing is

required. The performance of this testing is

intended to improve block valve availability

and thereby assure the capability of certain

accident mitigation strategies identified

within Abnormal and Emergency Operating

Procedures. Therefore, the requested revision

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. The requested revision to our previous

TS change request does not involve a

significant reduction in the margin of safety.

The requested revision is intended to help

assure block valve availability to support

certain accident mitigation strategies. This

additional assurance of block valve

availability and functioning increases the

margin of safety. Accordingly, the requested

revision will not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Hartsville Memorial Library,

147 West College Avenue, Hartsville,

South Carolina 29550.

Attorney for licensee: R. E. Jones,

General Counsel, Carolina Power &

Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William H.

Bateman.

Commonwealth Edison Company,

Docket Nos. 50–373 and 50–374, LaSalle

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of amendment request:

December 14, 1994.

Description of amendment request:

The proposed amendments would

revise technical specifications related to

allowed outage times (AOT) and

surveillance test intervals (STI) for

certain actuation instrumentation in the

reactor protection system (RPS), primary

containment isolation system (PCIS),

emergency core cooling system (ECCS),

recirculation pump trip, reactor core

isolation cooling (RCIC), control rod

withdrawal block, monitoring, and

feedwater/main turbine trip systems.

These changes are generally consistent

with General Electric topical reports

which have been reviewed and

approved by the NRC. The changes also

include revising the Feedwater/Main

Turbine Trip LCO 3.3.8 action statement

to achieve consistency with existing

instrumentation LCOs; deleting the

surveillance of the APRM Neutron

Flux—High, Setdown functional unit in

Operational Condition 1; revising the

applicability of the provisions of

Specification 4.0.4 to several Reactor

Protection System and Control Rod

Withdrawal Block Instrumentation

surveillance requirements; adding the

requirement to perform shiftly channel

checks for applicable RPS, PCIS, ECCS,

and RCIC instrumentation channels

equipped with master trip units; and

other changes to correct typographical

errors and to delete cycle specific

footnotes which are no longer

applicable.

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:

It has been determined that the changes do

not constitute a Significant Hazards

Consideration. Based on the criteria for

defining a significant hazards consideration

established in 10 CFR 50.92, operation of

LaSalle County Station Units 1 and 2 in

accordance with the proposed amendment

will not:

(1) Involve a significant increase in the

probability or consequences of an accident

previously evaluated because:

a. The proposed changes increase the STI

and AOT for actuation instrumentation

supporting RPS, ECCS, Isolation, CRBF,

RCIC, ATWS-RPT, EOC-RPT, Monitoring,

and Feedwater/Main Turbine Trip System

Actuation functions. There are no changes in

instrumentation configuration and function,

and no instrumentation setpoints are

changed. Because of this there is no change

in the probability of occurrence of an

accident or the consequences of an accident

or the consequences of malfunction of

equipment. With respect to the probability of

equipment malfunction, topical reports

prepared by GE demonstrate that there is a

reduction in scram frequency for the RPS, but

in the case of the ECCS there is a small

increase in the unavailability of the water

injection function. This increase in

unavailability was judged acceptable by GE.

The NRC concurred with this conclusion in

its review and approval of the topical reports.

The proposed changes are consistent with the

Safety Evaluation Reports issued for the

topical reports.

b. The changes proposed for the

Feedwater/Main Turbine Trip LCO action

statements provide actions which are

consistent with presently existing

instrumentation LCOs. The design and

function of the feedwater/main turbine trip

instrumentation to trip the feedwater pumps

and the main turbine upon detection of a

Level 8 event is not altered. The probability

and/or consequences of this moderate

frequency transient are not increased.

c. The APRM Neutron Flux—High,

Setdown scram setting provides adequate

thermal margin between the setpoint and the

safety limits for operation at low pressure

and low flow during a plant startup. This

function remains in effect until the mode

switch is placed in the Run (Operational

Condition 1) position, at which time it is

bypassed. Deleting the requirement for the

surveillance of the APRM Neutron Flux—

High, Setdown functional unit in Operational

Condition 1 is appropriate since its function

is not applicable in this mode. This deletion

serves to achieve consistency between

Technical Specification Tables and the Bases

section.

d. The changes associated with

Specification 4.0.4 are administrative in

nature and are intended to provide the plant

operators with better guidance for its

application. In cases where complete

surveillances cannot be achieved, such as

during a plant shutdown, then the required

surveillances will be performed within 24

hours of entering the Mode or condition in

which the surveillance is required. The

stabilization of the plant will be of primary

consideration. This change does not affect

the evaluation for any accident presented in

Chapter 15 of the UFSAR. The APRM Fixed

Neutron Flux—High quarterly functional

tests most of the APRM channel equipment

associated with the APRM Neutron Flux—

High, Setdown scram.

Additionally, the expected result of the

functional tests associated with the SRMs,

IRMs, and APRMs is to demonstrate the

operability of the instrumentation. Therefore,

24 hours is a reasonable time to permit the

surveillances to be performed upon entering

the mode or condition in which the

surveillance is required.

e. The proposal to include the performance

of channel checks as requirements of

technical specifications is administrative in

nature. Presently, channel checks performed

for the applicable analog instrumentation in

reactor vessel water level applications is

controlled solely by procedure. Adding this
Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices 11129

requirement to the technical specifications

provides for the appropriate controls of the

surveillances, above and beyond that

presently controlled by procedure.

f. The proposed administrative changes are

offered to correct typographical errors and

delete cycle specific footnotes which are no

longer applicable. The nature of the changes

precludes them from impacting previously

analyzed accidents.

The proposed changes therefore do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

(2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated because:

a. The proposed changes increase the STI

and AOT for certain actuation

instrumentation in the RPS, ECCS, Isolation,

CRBF, RCIC, ATWS-RPT, EOC-RPT,

Monitoring, and Feedwater/Main Turbine

Trip systems. There are no changes in

instrumentation configuration and function,

and no instrumentation setpoints are

changed.

b. The changes to the Feedwater/Main

Turbine Trip LCO action statements allow

the plant operators a maximum degree of

operational flexibility, while maintaining the

instrumentation and protection needed for

terminating the feedwater controller failure

transient. The single failure proof criterion of

the level sensors is maintained, and the logic

of the protective instrumentation is not

compromised. The changes to the LCO action

statements do not constitute a change to the

facility or its operation as described in the

Safety Analysis Report.

c. Deleting the requirement for surveilling

the APRM Neutron Flux—High, Setdown

functional unit in Operating Condition 1

does not degrade thermal margins. The

margin accommodates the anticipated

maneuvers associated with plant power

ascension. During a plant shutdown, rod

insertion maneuvers, recirculation flow

reduction, and xenon build-in all contribute

to negative reactivity insertion which

precludes the degradation and violation of

thermal margins. The functions of the

APRMs required to be OPERABLE in

Operational Condition 1 which are in effect

remain to ensure that reactor core thermal

margins are not compromised.

d. The conduct of neutron instrument

functional tests in the plant mode or

condition in which the trips are applicable

eliminates unnecessary testing during normal

plant operations. The expected result of the

functional testing is to demonstrate the

operability of the instruments. The failure of

any single instrument channel will neither

cause nor prevent either a reactor scram or

a control rod block.

e. Including the performance of channel

checks for the applicable analog

instrumentation as part of the technical

specifications transfers control of the

required surveillances from procedure to the

technical specifications, as appropriate. The

administrative nature of this change does not

alter the functions, setpoints, or

configuration of the associated

instrumentation.

f. The administrative nature of the changes

prevents them from affecting the functions,

setpoints, or configuration of the associated

instrumentation from being affected by the

changes.

The proposed changes do not create the

possibility for an accident or malfunction of

a different type than any previously

evaluated in the UFSAR.

(3) Involve a significant reduction in the

margin of safety because:

a. Setpoints are based upon the drift

occurring during an 18 month calibration

interval. The bases in the Technical

Specifications either do not discuss STI, or

state ‘‘* * * one channel may be inoperable

for brief intervals to conduct required

surveillance.’’ The proposed changes are

bounded by the analyses of the topical

reports. These analyses, which were prepared

by GE and approved by the NRC, examined

the effects of extending STI and AOT and

found that the proposed changes would not

involve a significant reduction in the margin

of safety.

b. The proposed changes to the turbine trip

LCO action statements do not change any of

the settings of the Level 8 setpoints. The

single failure criteria of the multiple level

sensors which sense and detect the Level 8

setpoint remains intact. The LCO maintains

the requirement that no single instrument

failure will prevent the feedwater pump

turbines and main turbine trip on a valid

Level 8 signal. Scram trip signals from the

turbine retain the design feature that a single

failure will neither initiate nor impede the

initiation of a reactor scram (trip).

c. The setting, function, and conditional

requirements of the APRM Neutron Flux—

High, Setdown function are not altered. This

change serves to achieve consistency

between two Technical Specifications Tables.

This eliminates the need for surveilling a

function in a mode which is not applicable.

The functions of the APRMs required to be

OPERABLE in Operational Condition 1

remain to ensure that reactor core thermal

margins are not compromised.

d. The reference to 4.0.4 applicability will

assist to ensure consistent interpretation of

the technical specifications by the plant

operators. This assists in ensuring that the

plant is operated within technical

specification limitations. This change does

not affect trip instrumentation setpoints, and

the scram function of the RPS is assured by

the weekly functional testing of the Manual

Scram.

e. Including the instrumentation channel

checks as part of technical specification

requirements provides an appropriately

regimented method of controlling the

conduct of the surveillances. None of the

functions, setpoints, or configuration of the

associated analog instrumentation is affected

by this administrative change.

f. The administrative nature of the changes

serves to provide more concise guidance to

the plant operating staff, and as such do not

impact the safety margin.

The proposed changes do not significantly

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: Public Library of Illinois

Valley Community College, Rural Route

No. 1, Oglesby, Illinois 61348

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60690.

NRC Project Director: Robert A. Capra.

Commonwealth Edison Company,

Docket Nos. 50–373 and 50–374, LaSalle

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of amendment request: January

13, 1995.

Description of amendment request:

The proposed amendments would

revise the pressure alarm setpoint

allowable values for the emergency core

cooling system (ECCS) and reactor core

isolation cooling (RCIC) system ‘‘keep

filled’’ pressure instrumentation

channels. The purpose of the proposed

change is to lower the setpoint

allowable values for these parameters to

more realistic values based upon

calculations performed by the licensee

reflecting design changes and system

performance. Also, the term ‘‘setpoint’’

is being changed to ‘‘setpoint allowable

value’’ to clarify the use of the values.

Additionally, two administrative/

editorial changes are included to delete

technical specification footnotes which

are no longer applicable.

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:

Commonwealth Edison has evaluated the

proposed Technical Specification

Amendment and determined that it does not

represent a significant hazards consideration.

Based on the criteria for defining a significant

hazards consideration established in 10 CFR

50.92, operation of LaSalle County Station

Units 1 and 2 in accordance with the

proposed amendment will not:

(1) Involve a significant increase in the

probability or consequences of an accident

previously evaluated because:

a. The proposed change in the technical

specification allowable values for the ECCS

and RCIC discharge line ‘‘keep filled’’ alarm

instrument channels does not change the

design bases or function of these systems as

described in the technical specifications and

UFSAR. An analysis performed by

engineering demonstrates that the proposed

allowable values are sufficient for verifying

that the ECCS and RCIC pump discharge

lines are full of water. In addition, setpoint
11130 Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices

calculations have been performed to verify

that sufficient margin exists between the

recommended calibration setpoints and the

analytical limits for these instrument

channels to account for all applicable

instrument errors. This provides high

assurance that the trip setpoints of these

instrument channels will not drop below the

minimum required value. The ‘‘keep filled’’

instrumentation is not a factor in the

assumptions of any accidents, thus, the

probability of analyzed accidents is not

increased.

b. The proposed technical specification

amendment does not revise the configuration

of the ECCS and RCIC discharge line ‘‘keep

filled’’ instrument channels or sensing lines.

The proposed setpoint allowable values and

associated calibration setpoints are within

the calibration ranges of the existing pressure

switches. Thus, implementation of the

proposed amendment does not involve any

physical alterations to the plant except for

the recalibration of the pressure switches to

the new calibration setpoints.

c. The ECCS and RCIC discharge line ‘‘keep

filled’’ instrument channels only perform a

monitoring function. Other than ensuring

system readiness they do not perform a

function important to safety. Thus, the

probability of a ECCS or RCIC failure is not

increased since the operation and function of

the ECCS and RCIC discharge line fill

systems is not affected by this change.

d. The failure of a ECCS or RCIC discharge

line fill system will not go undetected by the

proposed change, since water leg pump trips

are annunciated in the control room. In

addition, quarterly surveillances are

performed on these pumps to check for

degradation.

e. The ECCS and RCIC discharge line fill

systems are not used to mitigate the

consequences of an accident or transient.

These systems are not required after the

ECCS and RCIC pumps are activated.

Therefore, the proposed change does not

cause an increase in the probability or

consequences of an accident previously

evaluated.

(2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated because: This technical

specification amendment only lowers the trip

setpoint allowable values for the ECCS and

RCIC discharge line ‘‘keep filled’’ alarm

instrumentation channels. As described

above, the proposed setpoint allowable

values are sufficient for verifying that the

ECCS and RCIC discharge lines are full of

water. Thus, the probability of a water

hammer occurring during system activation

for a surveillance test is not increased. In

addition, each instrument channel is

independent from the other channels so that

a failure in one channel will not propagate

to another channel. Therefore, the operation

of the facility in accordance with the

proposed amendment does not create the

possibility of a new or different kind of

accident.

(3) Involve a significant reduction in the

margin of safety because: The margin of

safety is not affected by this amendment,

because this change involves monitoring

instrumentation only. The purpose of the

ECCS and RCIC discharge line ‘‘keep filled’’

alarms is to alert the operators when a ECCS

or RCIC system may not be operable due to

empty or partially empty discharge lines. The

proposed amendment does not alter or

degrade this function, since the new setpoint

allowable values are adequate for verifying

that the discharge lines are full of water.

Therefore the operation of the facility in

accordance with the proposed amendment

does not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: Public Library of Illinois

Valley Community College, Rural Route

No. 1, Oglesby, Illinois 61348.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60690.

NRC Project Director: Robert A. Capra.

Consumers Power Company, Docket No.

50–255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request: January

13, 1995

Description of amendment request:

The proposed amendment would

modify the required settings, and

allowable ‘‘as found’’ and ‘‘as left’’

tolerances for the primary and

secondary safety valves. The proposed

limits would allow installed primary

and secondary valve settings to be

within a 3% tolerance of their nominal

settings, but would require returning the

valve settings to within 1% of the

nominal settings if the valves are

removed from the piping for

maintenance or testing.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The following evaluation supports the

finding that operation of the facility in

accordance with the proposed technical

specification change would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change to the Technical

Specifications increases the acceptable as

found tolerance for the pressurizer safety

valves. The most limiting overpressure event,

loss of external load, has been analyzed to

account for this change. The loss of external

load analysis was performed using a

conservative 25% steam generator tube

plugging and an initial pressurizer level of

67.8% (providing an approximate 10%

conservative margin above programmed

pressurizer level for full power). Primary and

secondary safety valve accumulation was

conservatively accounted for and the setpoint

tolerance of +3% was assumed. Reactor trip

on turbine trip was assumed to be disabled

and the atmospheric dump valves were

assumed unavailable. The results of the

analysis demonstrated primary and

secondary system pressures within 110% of

design pressures. Therefore, the

consequences of overpressurization events

will not be significantly increased with a

+3% tolerance on the primary safety valve

setpoints. The proposed Technical

Specifications change will not affect normal

plant operation and will not increase the

probability of an accident.

A review of all DNB [departure from

nucleate boiling] analyses was performed to

ensure that predicted pressurizer pressures

for those analyses would not be affected by

a -3% tolerance on the lowest setpoint valve.

The DNB analyses for which significant

primary system pressure increases were

predicted do not result in pressures high

enough to lift the pressurizer safety valves

with the proposed tolerance. A conservative

DNB analysis that bounds the consequences

of inadvertent opening of a pressurizer safety

valve has also been previously performed

with predicted acceptable results. If a

pressurizer safety valve were to stick open,

the consequences would be bounded by the

small break LOCA [loss-of-coolant accident]

analysis. Therefore, the consequences due to

a -3% tolerance on the primary safety valve

setpoints will not increase the consequences

or probability of an accident.

The proposed revision removes the

requirement for one operable pressurizer

safety valve to be installed whenever the

reactor head is on the vessel. Instead,

proposed Specification 3.1.7.1 requires all

pressurizer safety valves to be operable above

cold shutdown, and overpressure protection

during cold shutdown is provided by existing

Specification 3.1.8.2, Power Operated Relief

Valves.

The proposed Technical Specifications

change also lists the lift settings for each of

the primary and secondary system safety

valves. This change will not affect the

operation or function of the valves.

Therefore, the probability and consequences

of previously evaluated accidents will not be

increased.

2. Create the possibility of a new or

different kind of accident from any

previously evaluated.

The proposed changes to Technical

Specifications will not affect the manner in

which the plant operates. The proposed

increase in pressurizer safety valve lift setting

tolerance could change the pressure at which

the valves open in an overpressurization

event, but would not create the possibility of

a new or different kind of accident. Since

Technical Specification 3.1.8 addresses

primary system overpressurization during

cold shutdown, the proposed removal of the

requirement for an operable pressurizer

safety valve to be installed whenever the

reactor head is on the vessel will not create
Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices 11131

the possibility of a new overpressurization

event during cold shutdown. The proposed

change to list the lift settings for the

individual primary and secondary safety

valves will have no effect on the safety

function of the valves. Therefore, the

proposed changes will not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The proposed changes to Technical

Specifications do not affect the DNB analyses

that have been previously performed. The

most limiting overpressurization event, loss

of external load, has been conservatively

analyzed accounting for the proposed

changes and demonstrated that the primary

and secondary system pressures remain

within 110% of the design pressures.

Overpressurization during cold shutdown is

addressed by Technical Specification 3.1.8.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Power Company,

212 West Michigan Avenue, Jackson,

Michigan 49201.

NRC Project Director: John N.

Hannon.

Consumers Power Company, Docket No.

50–255, Palisades Plant, Van Buren

County, Michigan.

Date of amendment request: February

10, 1995.

Description of amendment request:

The proposed amendment would

modify the Technical Specifications to

allow a one time deferral of several 18-

month interval surveillance tests until

the upcoming scheduled refueling

outage to avoid the necessity of

imposing a plant shutdown solely for

the sake of their performance.

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 following evaluation supports the

finding that operation of the facility in

accordance with the proposed Technical

Specifications (TS) would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Deferring surveillance testing will

introduce no new operating conditions,

change no equipment operating procedures,

and change no plant systems or equipment.

Therefore, operation of the facility in

accordance with the proposed TS would not

result in a significant increase in the

probability of an accident previously

evaluated.

Deferring surveillance testing of snubbers

and instrument channels could allow minor

degradations of snubber condition or small

changes in instrument setpoints or

calibration to progress some amount beyond

that point which would occur with a shorter

surveillance interval. A review of the recent

test history for the subject surveillance

indicates that no significant snubber

degradation or instrument drift was found. It

is not expected that, even with the proposed

surveillance deferral, snubber conditions or

instrument settings will be found to exceed

conditions allowable by the Technical

Specifications. Therefore, operation of the

facility in accordance with the proposed TS

would not result in a significant increase in

the consequences of an accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously evaluated.

Deferring surveillance testing will

introduce no new operating conditions,

change no equipment operating procedures,

and change no plant systems or equipment.

Therefore, operation of the facility in

accordance with the proposed TS would not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. Involve a significant reduction in a

margin of safety.

A review of past performance of the subject

surveillance tests indicate that the requested

deferral of testing would not have a

significant effect on the results of the tests

when they are performed prior to the startup

for cycle 12. Most of the affected

instrumentation is monitored each shift by

channel checks, which would disclose major

failures or significant drift. Therefore,

operation of the facility in accordance with

the proposed TS would 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: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Power Company,

212 West Michigan Avenue, Jackson,

Michigan 49201.

NRC Project Director: John N.

Hannon.

Duke Power Company, Docket Nos. 50–

369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of amendment request:

November 2, 1994.

Description of amendment request:

The proposed amendment would delete

the content of the Appendix B,

Environmental Protection Plan (EPP)

and modify License Conditions 2.C.(2)

to delete that portion which refers to the

EPP. Specifically, the requirements for

non-radiological environmental

monitoring have been completed. The

radiological environmental monitoring

requirements have been incorporated

into Appendix A (the Technical

Specifications). There would be no

impact on the continued safety of the

McGuire station by deleting Appendix

B.

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:

Deletion of the Environmental Protection

Plan and modifying License Condition 2.C.(2)

will have no impact on the probability or

consequences of an accident previously

evaluated because the changes will not have

any impact upon the design or operation of

any plant systems or components.

The proposed revision will not create the

possibility of a new or different kind of

accident from any previously evaluated

because the revision is administrative in

nature and will not change the types and

amounts of effluent that will be released.

The proposed revision will not reduce a

margin of safety because it is administrative

in nature and will not effect the margin of

safety as defined in the basis for any

Technical Specifications.

Accordingly, this proposed changes does

not involve a significant hazard.

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: Atkins Library, University of

North Carolina, Charlotte (UNCC

Station), North Carolina 28223.

Attorney for licensee: Mr. Albert Carr,

Duke Power Company, 422 South

Church Street, Charlotte, North Carolina

28242.

NRC Project Director: Herbert N.

Berkow.
11132 Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices

Duke Power Company, Docket Nos. 50–

369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of amendment request: January

18, 1995.

Description of amendment request:

The proposed amendments would

relocate the requirements for the seismic

instrumentation, meteorological

instrumentation, and loose-part

detection system from the Technical

Specifications to the Selected Licensee

Commitment (SCL) Manual. This will

allow future changes to these controls to

be performed under the provisions of 10

CFR 50.59. No changes are being made

to the technical content of the affected

Technical Specification pages.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Criterion 1

The requested amendments will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. Relocation of the

affected TS sections to the SLC Manual will

have no effect on the probability of any

accident occurring. In addition, the

consequences of an accident will not be

impacted since the above instrumentation

will continue to be utilized in the same

manner as before. No impact on the plant

response to accidents will be created.

Criterion 2

The requested amendments will not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. No new accident causal

mechanisms will be created as a result of

relocating the affected TS requirements to the

SLC Manual. Plant operation will not be

affected by the proposed amendments and no

new failure modes will be created.

Criterion 3

The requested amendments will not

involve a significant reduction in a margin of

safety. No impact upon any plant safety

margins will be created. Relocation of the

affected TS requirements to the SLC Manual

is consistent with the content of the

Westinghouse RSTS [Revised Standard

Technical Specifications], as the NRC did not

require technical specification controls for

the affected instrumentation in the RSTS.

The proposed amendments are consistent

with the NRC philosophy of encouraging

utilities to propose amendments that are

consistent with the content of the RSTS.

Based upon the preceding analyses, Duke

Power Company concludes that the requested

amendments do not involve a significant

hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Atkins Library, University of

North Carolina, Charlotte (UNCC

Station), North Carolina 28223.

Attorney for licensee: Mr. Albert Carr,

Duke Power Company, 422 South

Church Street, Charlotte, North Carolina

28242.

NRC Project Director: Herbert N.

Berkow.

Duke Power Company, Docket Nos. 50–

369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of amendment request: January

18, 1995.

Description of amendment request:

The amendments would revise

Technical Specification Table 4.3–3 to

allow the analog channel operational

test interval for radiation monitoring

instrumentation to be increased from

monthly to quarterly. The proposed

amendment changes would be

consistent with the guidance in Generic

Letter 93–05, ‘‘Line-Item Technical

Specifications Improvements to Reduce

Surveillance Requirements for Testing

During Power 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:

Criterion 1

The requested amendments will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. Decreasing the

frequency of the radiation monitor analog

channel operational test from monthly to

quarterly will have no impact upon the

probability or any accident, since the

radiation monitors are not accident initiating

equipment. Analysis of the previous test data

* * * shows that no significant degradation

of performance is to be expected by the

decrease in frequency. Therefore, the

requested amendments will have no adverse

impact upon the consequences of any

accident.

Criterion 2

The requested amendments will not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. As stated above, the radiation

monitors are not accident initiating

equipment. No new failure modes can be

created from an accident standpoint. The

plant will not be operated in a different

manner.

Criterion 3

The requested amendments will not

involve a significant reduction in a margin of

safety. Plant safety margins will be

unaffected by the proposed changes. No

safety equipment which is taken credit for in

accident analyses will be affected by the

requested amendments. The availability of

the affected radiation monitors will be

increased as a result of the proposed

amendments because the monitors will not

have to be made unavailable for testing as

frequently. In addition, radiation monitor

operating experience supports the proposed

amendments. Finally, the proposed

amendments are consistent with the NRC

position and guidance set forth in NUREG–

1366 and Generic Letter 93–05.

Based upon the preceding analyses, Duke

Power Company concludes that the requested

amendments do not involve a significant

hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Atkins Library, University of

North Carolina, Charlotte (UNCC

Station), North Carolina 28223.

Attorney for licensee: Mr. Albert Carr,

Duke Power Company, 422 South

Church Street, Charlotte, North Carolina

28242.

NRC Project Director: Herbert N.

Berkow.

Florida Power and Light Company, et

al., Docket Nos. 50–335 and 50–389, St.

Lucie Plant, Unit Nos. 1 and 2, St. Lucie

County, Florida

Date of amendment request: January

20, 1995.

Description of amendment request:

The proposed amendments will relocate

the operability requirements for the

INCORE DETECTORS (TS 3/4.3.3.2) to

the Updated Final Safety Analysis

Report, and revise Linear Heat Rate

surveillance 4.2.1.4, and Special Test

Exceptions surveillances 4.10.2.2,

4.10.4.2 (Unit 2 only), and 4.10.5.2,

accordingly.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes are administrative

in nature in that the specifications for
Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices 11133

operation and surveillance of the Incore

Instrumentation (ICI) System will be

relocated from the Technical Specifications

to the Updated Final Safety Analysis Report

for St. Lucie Unit 1 and Unit 2. Changes to

the system will be controlled by 10 CFR

50.59, and the safety analysis report is

required to be updated pursuant to 10 CFR

50.71(e). Relocation of these requirements to

the UFSAR is consistent with the NRC ‘‘Final

Policy Statement on Technical Specifications

Improvements for Nuclear Power Reactors’’

published in the Federal Register (58 FR

39132) dated July 22, 1993.

Incore instrumentation is not an accident

initiator nor a part of the success path(s)

which function to mitigate accidents

evaluated in the plant safety analyses. The

proposed technical specification change does

not involve any change to the configuration

or method of operation of any plant

equipment that is used to mitigate the

consequences of an accident, nor do the

changes alter any assumptions or conditions

in any of the plant accident analyses.

Therefore, 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.

(2) Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendment to relocate the

existing Technical Specification

requirements for the Incore Instrumentation

System to the Updated Final Safety Analysis

Report will not change the physical plant or

the modes of plant operation defined in the

Facility License. The change does not involve

the addition or modification of equipment

nor does it alter the design or operation of

plant systems. Therefore, 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.

(3) Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The proposed changes are administrative

in nature in that operating and surveillance

requirements for the Incore Instrumentation

System will be relocated from the Technical

Specifications to the Updated Final Safety

Analysis Report for St. Lucie Unit 1 and Unit

2. The ICI system is not used to actuate

safety-related equipment, provide interlocks,

or otherwise perform automatic plant control

functions. The system is used to monitor core

power distribution parameters whose limits

do involve a margin of safety; however, the

ICI system itself makes no contribution to

that margin of safety, and the power

distribution limits will not be changed by the

proposed amendment. Therefore, operation

of the facility in accordance with the

proposed amendment would not involve a

significant reduction in a margin of safety.

Based on the above discussion and the

supporting Evaluation of Technical

Specification changes, FPL has determined

that the proposed license amendment

involves no significant hazards

consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954–9003.

Attorney for licensee: Harold F. Reis,

Esquire, Newman and Holtzinger, 1615

L Street, NW., Washington, DC 20036.

NRC Project Director: David B.

Matthews.

Florida Power and Light Company,

Docket Nos. 50–250 and 50–251, Turkey

Point Plant Units 3 and 4, Dade County,

Florida

Date of amendment request: January

17, 1995.

Description of amendment request:

The licensee proposes to revise the

technical specifications to reference

Topical Report NF-TR–95–01 as the

documentation of the licensee’s

proficiency in performing certain reload

design calculations once the NRC has

evaluated and approved NR–TR–95–01.

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 amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The addition of the reference to FPL

[Florida Power and Light Company] topical

report which demonstrates FPL’s ability to

perform certain reload design calculations for

Turkey Point Units 3 and 4 is administrative

in nature and has no impact on the

probability or consequences of any Design

Bases Event (DBE) occurrences previously

evaluated. The reload design calculations

will be performed using methodologies and

computer codes approved by the NRC and

poses no increase in the probability or

consequences of any accident previously

evaluated.

The Core Operating Limits Report (COLR)

parameters will be evaluated every cycle to

ensure proper compliance with the Updated

Final Safety Analysis Report (UFSAR). These

limits will be evaluated in accordance with

10 CFR [Section] 50.59, which ensures that

the reload will not involve an increase in the

probability of occurrences or consequences of

an accident previously evaluated. Title 10

CFR [Section] 50.59 (2) states that a proposed

change involves an unreviewed safety

question (i) if the probability of occurrence

or the consequences of an accident or

malfunction of equipment important to safety

previously evaluated in the safety analysis

report may be increased. Consequently, since

any change to the reload core design analysis

must be evaluated relative to the more

restrictive evaluation criterion of 10 CFR

[Section] 50.59, then operation of the facility

in accordance with the proposed

amendments would not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The addition of the reference to FPL

topical report which demonstrates FPL’s

ability to perform certain reload design

calculations for Turkey Point Units 3 and 4

is administrative in nature and has no

impact, nor does it contribute in any way to

the possibility of a new or different kind of

accident from any accident previously

evaluated. No new accident scenarios, failure

mechanisms or limiting single failure events

are introduced as a result of the proposed

change.

The generation of the Axial Flux

Difference, Rod Bank Insertion limits and

K(Z) curve will be performed using NRCapproved methodology and are submitted to

the NRC, as a revision to the COLR, to allow

the NRC staff to trend. The Technical

Specifications will continue to require

operation within the core operating limits

and appropriate actions will be taken if these

limits are exceeded.

Title 10 CFR [Section] 50.59 permits a

licensee to make changes in the facility as

described in the safety analysis report

without prior Commission approval,

provided that the proposed changes does not

involve an unreviewed safety question. 10

CFR [Section] 50.59 (2) states that a proposed

change involves an unreviewed safety

question (ii) if a possibility for an accident

or malfunction of a different type than any

evaluated previously in the safety analysis

report may be created. Consequently, since

any change to the reload core design analysis

must be evaluated relative to the more

restrictive evaluation criterion of 10 CFR

[Section] 50.59, then operation of the facility

in accordance with the proposed

amendments 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 amendments would not

involve a significant reduction in a margin of

safety.

The margin of safety is not affected by FPL

performing the reload design calculations for

Turkey Point Units 3 and 4. The supporting

Technical Specification values are defined by

the accident analyses which are performed to

conservatively bound the operating

conditions defined by the Technical

Specifications. The development of the limits

for future reloads will continue to conform to

the methodology described in NRC approved

documentation. In addition, each future

reload will involve a 10 CFR [Section] 50.59
11134 Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices

review to assure that operation of the units

within the cycle specific limits will not

involve a reduction in a margin of safety. 10

CFR [Section] 50.59 (2) states that a proposed

change involves an unreviewed safety

question (iii) if the margin of safety as

defined in the basis for any technical

specification is reduced. Consequently, since

any change to the reload core design analysis

must be evaluated relative to the more

restrictive evaluation criterion of 10 CFR

[Section] 50.59, then operation of the facility

in accordance with the proposed

amendments would not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration. The NRC staff, however,

considers that the licensee’s statements

relative to 10 CFR Section 50.59

evaluations to be performed in the

future are not relevant to the proposed

no significant hazards determination.

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199.

Attorney for licensee: Harold F. Reis,

Esquire, Newman and Holtzer, P.C.,

1615 L Street, NW., Washington, DC

20036.

NRC Project Director: David B.

Matthews.

IES Utilities Inc., Docket No. 50–331,

Duane Arnold Energy Center, Linn

County, Iowa

Date of amendment request: October

28, 1994.

Description of amendment request:

The proposed amendment revises the

Duane Arnold Energy Center (DAEC)

Operating License by deleting a

condition of the license that requires a

‘‘Plan for Integrating Scheduling of

Plant Modifications for the Duane

Arnold Energy Center’’ (the Plan).

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 provided below:

(1) The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. No physical changes to

the facility will occur as a result of this

amendment. Work activities will continue to

receive the appropriate level of review in

accordance with DAEC procedures and

practices. The organizational structure that

controls and manages these activities remains

unchanged and will assure that activities are

prioritized and performed in a manner

consistent with plant safety. The proposed

amendment removes an administrative

burden that is no longer required.

(2) The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated. No changes to the

physical design and/or operation of the plant

will occur as a result of this amendment. The

processes by which activities are planned,

prioritized, and controlled are not affected.

The appropriate level of technical review and

management oversight continue to be

performed in accordance with existing

procedures and practices to assure that

activities are performed in a manner

consistent with plant safety.

(3) The proposed amendment does not

involve a significant reduction in a margin of

safety. As stated earlier, no changes to the

physical design and/or operation of any plant

systems will occur as a result of this

amendment. Work activities will continue to

receive the appropriate technical review and

management oversight to assure that

activities are prioritized and performed in a

manner consistent with plant safety. The

amendment removes an administrative

burden that is no longer required.

Based on the above, we have determined

that the proposed amendment will not

involve a significant hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids,

Iowa 52401.

Attorney for licensee: Jack Newman,

Kathleen H. Shea, Morgan, Lewis &

Bouckins, 1800 M Street NW.,

Washington, DC 20036.

NRC Project Director: Leif J.

Norrholm.

Niagara Mohawk Power Corporation,

Docket No. 50–220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego

County, New York

Date of amendment request: January

24, 1995.

Description of amendment request:

The proposed amendment would revise

Technical Specification 3.4.1, ‘‘Leakage

Rate,’’ to reduce the allowable leakage

rate of the reactor building from 2000

cubic feet per minute (cfm) to 1600 cfm.

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 operation of Nine Mile Point Unit 1,

in accordance with the proposed

amendment, will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

Secondary containment and RBEVS

[Reactor Building Emergency Ventilation

System] are not initiators or precursors to an

accident. Secondary containment provides a

pressure boundary, with limited in-leakage,

for the purpose of preventing a ground level

unfiltered release of radioactivity. RBEVS

responds to accidents involving release of

radioactivity to the secondary containment

by maintaining a negative pressure inside

secondary containment and by providing an

elevated release. Therefore, a change to the

Reactor Building leakage rate cannot affect

the probability of an accident previously

evaluated.

Although the proposed change reduces the

Reactor Building leakage rate from 2000 cfm

to 1600 cfm consistent with system design,

there is no effect on the radiological

consequences of any previously analyzed

accident since the radiological analysis does

not assume exfiltration. Therefore, the

Technical Specification change does not

significantly increase the consequences of a

previously evaluated accident.

The operation of Nine Mile Point Unit 1,

in accordance with the proposed

amendment, will not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

The proposed change to the Reactor

Building leakage rate from 2000 cfm to 1600

cfm does not involve any accident precursors

or initiators. During an accident involving a

release of radioactivity to the secondary

containment, the RBEVS would be operable

and provide filtration of containment

atmosphere prior to release to the

environment. This change does not involve

any physical modifications to the system,

thus the system will operate as designed.

Therefore, the proposed Technical

Specification change will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The operation of Nine Mile Point Unit 1,

in accordance with the proposed

amendment, will not involve a significant

reduction in a margin of safety.

The proposed change in Reactor Building

in-leakage from 2000 cfm to 1600 cfm in

Specification 3.4.1 and the associated basis is

to be consistent with system design and

reflect the leakage rate associated with

approximately one building air volume

change per day. The resulting accident

analysis remains unchanged since the

radiological analysis does not assume any

exfiltration. Therefore, the proposed change

will not involve a significant reduction in the

margin of safety as defined in the basis for

any Technical Specification.

Therefore, as determined by the above

analysis, this proposed amendment involves

no significant hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request
Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices 11135

involves no significant hazards

consideration.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Project Director: Ledyard B.

Marsh.

Niagara Mohawk Power Corporation,

Docket No. 50–220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego

County, New York

Date of amendment request: February

1, 1995.

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 3.6.13,

‘‘Remote Shutdown Panels.’’ TS 3.6.13

currently requires that if the valve

controls or monitoring instrumentation

on the Remote Shutdown Panels are

inoperable, they must be restored to an

operable status within 24 hours or the

plant shall be shut down. The proposed

change would require inoperable valve

control functions be restored to an

operable status within 30 days or the

plant shall be shut down. The proposed

change would also specify that required

inoperable monitoring instrumentation

functions be restored to an operable

status within 30 days or that an

alternate method of monitoring the

parameter be established within 30 days

and the required function be restored to

an operable status within 90 days or the

plant shall be shut down.

The proposed amendment would also

make minor editorial changes to TS

Table 3.6.13–1 so that the table entries

would be consistent with the proposed

revisions to TS 3.6.13.

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 operation of Nine Mile Point Unit 1,

in accordance with the proposed

amendment, will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The remote shutdown panel monitoring

instruments and controls are not initiators or

precursors to an accident. The remote

shutdown panels provide the operator with

sufficient monitoring instruments and

controls to place and maintain the plant in

a safe shutdown condition from a location

other than the control room. Therefore, the

proposed changes to Specification 3.6.13,

‘‘Remote Shutdown Panels,’’ cannot affect

the probability of a previously evaluated

accident.

The proposed changes, in part, require that

one channel (on either panel) for each

function be operable. This change could

potentially avoid an unnecessary plant

shutdown without affecting an operator’s

ability to cope with a control room

evacuation. One channel of each function is

adequate to assure a safe shutdown. The

proposed changes would also allow 30 days

to restore an inoperable function to an

operable status. As indicated in the ITS

[Improved Standard Technical

Specifications], the allowed time of 30 days

is acceptable based on operating experience

and the low probability of an event that

would require evacuation of the control

room. With one or more monitoring

instrument functions inoperable, the

proposed change gives an operator an

additional option. Specifically, the operator

is allowed 30 days to establish an alternate

method of monitoring the parameter and 90

days to restore the function to operable

status. The use of an alternate method is

acceptable since it will provide the operator

with indication of the parameter of interest.

The remote shutdown panels will not be

required to be operable in hot shutdown

because the plant is already subcritical and

in a condition of reduced reactor coolant

inventory energy. Because this Specification

no longer applies to hot shutdown and to be

consistent with the guidance provided in the

ITS, Specification 3.6.13.d will require that

the plant be brought to a hot shutdown

condition (versus cold shutdown condition)

in 12 hours. As indicated in the ITS, the 12-

hour completion time is reasonable based on

operating experience. The Bases Section to

3.6.13 and 4.6.13 was revised to be consistent

with the proposed changes to the

Specification. The Bases currently indicates

that one remote shutdown panel is required

to be operable. As explained above, one

channel of each required function is required

to maintain remote shutdown operability. In

summary, the proposed changes will not

affect the ability of the Remote Shutdown

System to provide the operator with

sufficient instrumentation and controls to

place and maintain the plant in a safe

shutdown condition from a location other

than the control room. Therefore, the

consequences of an event requiring a control

room evacuation will not significantly

increase.

Editorial changes were made to Table

3.6.13–1 to be consistent with the changes

made to the Specification. Specifically, the

word ‘‘INSTRUMENT’’ was changed to

‘‘FUNCTION’’ and the words ‘‘PANEL

MONITORING’’ were changed to the words

‘‘PANELS FUNCTIONS.’’ These changes

make it clear that one channel of each

function, on either panel is acceptable to

maintain operability. The emergency

condenser condensate return valve control

and motor-operated steam supply valves

control were relocated from Specification

3.6.13.b to Table 3.6.13–1 to be consistent

with the proposed changes.

Based on the above, the consequences of an

accident previously evaluated are not

significantly increased.

The operation of Nine Mile Point Unit 1,

in accordance with the proposed

amendment, will not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

The changes do not introduce any new

accident precursors and do not involve any

alterations to plant configurations which

could initiate a new or different kind of

accident. The proposed changes require that

one channel of each function be operable to

assure the remote shutdown panels can meet

their intended function. No changes have

been made which will affect the operation of

the remote shutdown panels in a way which

would create a new or different kind of

accident. Therefore, the proposed changes

will not create the possibility of a new or

different kind of accident from any

previously evaluated.

The operation of Nine Mile Point Unit 1,

in accordance with the proposed

amendment, will not involve a significant

reduction in a margin of safety.

The proposed changes will not affect the

ability of the Remote Shutdown System to

provide the operator with sufficient

instrumentation and controls to place and

maintain the plant in a safe shutdown

condition from a location other than the

control room. The ability to respond to a

control room evacuation is maintained with

one channel operable for each required

function. The allowed outage time of 30 days

is acceptable based on operating experience

and the low probability of an event requiring

control room evacuation. Therefore, the

proposed changes do not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Project Director: Ledyard B.

Marsh.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of amendment request: January

10, 1995.

Description of amendment request:

The proposed amendment request

would revise Technical Specifications

by deleting the power range, neutron

flux, high negative rate trip from Tables

2.2–1, 3.3–1, and 4.3–1, and delete the

associated Bases Section 2.0.
11136 Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices

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 (SHC), which is presented

below:

* * * The proposed changes would not

involve an SHC because the changes would

not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The deletion of the power range, neutron

flux, high negative rate trip will not

adversely affect plant operations. As has been

presented and accepted by the NRC Staff in

previous docketed correspondence, the

dropped RCCA [rod cluster control assembly]

accident analysis does not rely on this trip

to safely shut down the plant. The safety

analysis of the plant is unaffected by the

proposed changes. Since the safety analysis

is unaffected, the calculated

radiologicalreleases associated with the

analysis are not affected. Therefore, the

proposed changes will not increase the

probability or consequences of an accident

previously evaluated.

The reactor trip system is used to mitigate

accidents. There have been instances, during

calibration of these units, where a single

channel has generated a trip signal. Leaving

this in place when it is not necessary could,

therefore, cause a reactor trip. The deletion

of one trip function will, therefore, slightly

decease, not increase, this probability.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The reactor trip system is used to mitigate

accidents, and the only way that it can

initiate an event is by causing the reactor to

trip when it is unnecessary. This possibility

of the generation of a false trip signal has

already been evaluated in the safety analysis.

This modification will physically remove or

disable the power range, neutron flux trip

and will therefore decrease the possibility for

the generation of a false trip signal.

Therefore, the proposed change cannot create

a new or different kind of accident from any

previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The proposed change which deletes the

power range, neutron flux, high negative rate

trip will have no impact on the margin of

safety. The current safety analysis for

Millstone Unit No. 3 does not credit this trip

for any events; therefore, removal of this trip

from the technical specifications will not

affect the margin of safety for any analyzed

events.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resource Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich, CT

06360.

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141–0270.

NRC Project Director: Phillip F.

McKee.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of amendment request: January

23, 1995.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS) by 1)

adding a new Section 3/4.5.5 which

provides a limiting condition for

operation, an action statement, a

surveillance requirement, and a

corresponding bases section, for the

trisodium phosphate (TSP) baskets

which will be installed in the next

refueling outage; 2) deleting Section 3/

4.6.2.3 and Bases 3/4.6.2.3 related to the

spray additive system which are no

longer needed since the chemical

addition tank is being abandoned; and

3) updating Index Pages viii, ix, and xiv

to reflect the above changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration (SHC), which is presented

below:

* * * The proposed changes do not

involve an SHC because the changes would

not:

1. Involve a Significant Increase in the

Probability or Consequences of an Accident

Previously Evaluated.

The plant change affects the chemical

composition of the QSS [quench spray

system] flow and the method of sump pH

control, which are important for containment

heat removal/pressure mitigation (MSLB and

LOCA) [main steamline break and loss-of

coolant accident] and fission product

removal (LOCA). However, this change does

not affect the probability of occurrence of

these accidents. Since the TSP baskets are

passive devices located inside the

containment, they cannot initiate a transient

or affect the probability of occurrence of any

previously evaluated accident.

The design change will not adversely affect

the radiological doses for the DBA [design

basis accident] LOCA at the Exclusion Area

Boundary, Low Population Zone, Millstone

Unit No. 3 Control Room, Millstone Unit No.

2 Control Room, and the Millstone Technical

Support Center. Also, the change will not

adversely affect the calculated peak clad

temperature (PCT) for the DBA LOCA.

2. Create the Possibility of a New or

Different Kind of Accident from any

Previously Analyzed.

The change does not create a malfunction

that is different from those previously

evaluated. The TSP baskets are passive

devices that have minimal impact on any

other systems except through water

chemistry. The change in water chemistry

does not adversely affect any safety systems.

The installation of the TSP baskets and the

abandonment of the CAT [chemical addition

tank] will not change the probability of a

malfunction of safety-related equipment.

Potential malfunctions relating to the TSP

powder, the 12 baskets which hold the TSP

powder, the QSS and other systems, and

equipment credited in the safety analysis

were evaluated and determined not to be

adversely affected by the change.

Additionally, the transient pH behavior of

the spray flow will not adversely affect

metals, coatings and elastomers in the

containment, and the performance of

associated safety functions is not affected.

Finally, the change in the chemical

composition of the QSS solution will not

affect the operability of this system or its

ability for containment heat removal and

pressure mitigation.

3. Involve a Significant Reduction in the

Margin of Safety.

The design changes do not adversely affect

the ability of the QSS to perform the function

of containment heat removal, pressure

mitigation and fission product (iodine)

retention. The design changes do not

adversely affect any equipment credited in

the safety analysis. Also, the design changes

to not increase the calculated peak clad

temperature (PCT) or the offsite doses due to

the design basis LOCA. Therefore, there is no

impact on the margin of safety as specified

in the technical specifications.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resource Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich, CT

06360.

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141–0270.

NRC Project Director: Phillip F.

McKee.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of amendment request: January

24, 1995.
Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices 11137

Description of amendment request:

The amendment request would revise

the Technical Specification Section

3.2.3.1.a and Table 2.2–1 to decrease the

acceptance criterion for measured

reactor coolant system (RCS) flow rate

from 387,480 gallons per minute (gpm)

to 371,920 gpm.

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 (SHC), which is presented

below:

* * * The proposed changes do not

involve an SHC because the changes would

not:

1. Involve a Significant Increase in the

Probability or Consequence of an Accident

Previously Evaluated.

An evaluation of the 4% decrease in the

RCS total flow rate limit has shown that the

change does not significantly impact the

design basis analyses. Therefore, the change

will not increase the consequences of an

accident previously evaluated.

There are no actual plant changes that will

result from this technical specification

change. Instead, the technical specification

requirement for minimum total RCS flow rate

is being changed to provide operational

benefit without compromising safety. Since

there are no plant changes, there is no effect

on the probability of occurrence of

previously evaluated accidents.

The change will have a negligible impact

on the small break loss of coolant accident

(LOCA) and large break LOCA analyses. The

PCT [peak cladding temperature] acceptance

criteria will continue to be met with the

assumption of a 4% reduction in RCS flow

rate.

For the steam generator tube rupture event,

both the FSAR [Final Safety Analysis Report]

offsite dose analysis and the margin of steam

generator (SG) overfill were evaluated. It was

determined that the 4% reduction in RCS

flow rate will not adversely affect the offsite

doses or the margin to SG overfill and,

therefore, the FSAR conclusions remain

unchanged.

In the evaluation of non-LOCA transients,

the DNB [departure from nucleate boiling] is

the most affected parameter due to a change

in flow rate. It was concluded that the 4%

reduction in RCS flow was acceptable and

there was margin to the DNB limit.

It is concluded that there is sufficient

margin to the system pressure, PCT and DNB

limits to offset the effect of the 4% flow rate

decrease and the calculated radiological

releases associated with the analysis are not

affected. Therefore, there is no effect on the

consequences of previously evaluated

accidents.

2. Create the Possibility of a New or

Different Kind of Accident from any

Previously Analyzed.

The low loop flow trip setpoint specified

in Technical Specification Table 2.2–1 is set

as a fraction of total flow. The flow fraction

is not being changed and no hardware

changes are required due to the reduction in

minimum flow. Also, the reduction in

minimum flow will not change the operation

of any plant equipment and it does not

modify plant operation.

Therefore, the reduction in minimum flow

does not introduce any new failure modes or

malfunctions and it does not create the

potential for a new unanalyzed accident.

3. Involve a Significant Reduction in the

Margin of Safety.

The proposed 4% decrease in the technical

specification limit for total RCS flow rate will

not adversely affect the results of the FSAR

accident analysis, and it is concluded that

this change is safe. The change does not

adversely affect any equipment credited in

the safety analysis, and it does not affect the

probability of occurrence of any plant

accident. Also, the change has a negligible

impact on the PCT, and it does not increase

the offsite doses or decrease the DNB below

its acceptance limit.

Therefore, the change does not have any

significant impact on the protective

boundaries, and there is no reduction in the

margin of safety as specified in the technical

specifications.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resource Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich, CT

06360.

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141–0270.

NRC Project Director: Phillip F.

McKee.

Omaha Public Power District, Docket

No. 50–285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: January

9, 1995.

Description of amendment request:

The proposed amendment to the

technical specifications (TSs) would

delete requirements for the toxic gas

monitoring system (TGMS) as contained

in TS 2.22 and TS 3.1, Table 3–3, item

29.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) The proposed changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The previously evaluated accidents

affected by this change are the on-site and

off-site toxic chemical releases. These events

have been re-evaluated for this proposed

change and have been shown to meet the

applicable regulatory screening criteria. The

deterministic analyses performed show that

the guidelines of Regulatory Guide 1.78 for

control room habitability are met for on-site

and most off-site chemicals. On-site chemical

sources originally present when the toxic gas

monitoring system was installed have been

removed from site or determined not to

exceed the deterministic analysis screening

requirements. For those off-site chemical

releases which did not meet the deterministic

screening criteria a probabilistic analysis was

performed. The probabilistic analysis

performed in support of this proposed

change shows that the probability of an offsite chemical release leading to 10 CFR 100

consequences is orders of magnitude less

than the SRP [Standard Review Plan] 2.2.3

guidelines. These results show that there is

no significant increase in the probability or

consequences of any accident previously

evaluated.

(2) The proposed changes do not create the

possibility of a new or different kind of

accident from any previously analyzed.

Only events involving chemicals for which

the TGMS provides an automatic detection/

isolation function are affected by this change.

As stated above, the potential events

involving these chemicals have been reevaluated using the appropriate regulatory

guidance and shown to satisfy either the

deterministic screening criteria of RG

[Regulatory Guide] 1.78, or to be

probabilistically insignificant compared to

the guidelines of SRP Section 2.2.3. These

results show that the proposed change will

not create the possibility of a new or different

kind of accident from any previously

evaluated. Therefore, the proposed change

does not create the possibility of a new or

different kind of accident from any

previously analyzed.

(3) The proposed changes do not involve

a significant reduction in a margin of safety.

The margin of safety is defined by the

regulatory basis for the existing TGMS,

namely NUREG–0737, Item III.D.3.4. The

analysis provided to support this proposed

change follows the regulatory guidelines of

RG 1.78 and SRP Section 2.2.3, as specified

in NUREG–0737, Item III.D.3.4. The analysis

shows that the applicable regulatory criteria

are met and the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102.

Attorney for licensee: LeBoeuf, Lamb,

Leiby, and MacRae, 1875 Connecticut
11138 Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices

Avenue, NW., Washington, DC 20009–

5728.

NRC Project Director: Theodore R.

Quay.

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of amendment requests:

February 6, 1995 (Reference LAR 95–

01).

Description of amendment requests:

The proposed amendments would

revise the combined Technical

Specifications (TS) for the Diablo

Canyon Power Plant, Unit Nos. 1 and 2,

to change TS 3/4.9.14.1, ‘‘Spent Fuel

Assembly Storage,’’ TS 3/4.9.14.2,

‘‘Spent Fuel Pool Boron Concentration,’’

TS 5.3.1, ‘‘Reactor Core—Fuel

Assemblies,’’ and TS 5.6.1, ‘‘Fuel

Storage—Criticality,’’ and add new TS

3/4.9.14.3, ‘‘Spent Fuel Assembly

Storage—Spent Fuel Pool Region 1.’’

The specific TS changes proposed are as

follows:

(1) The proposed changes to TS 3/

4.9.14 are:

(a) TS 3.9.14.1 and Figure 3.9–2

would be revised to allow the storage of

spent fuel assemblies with initial

enrichments up to 5.0 weight percent

uranium-235 (U–235) in Region 2 of the

spent fuel pool (SFP). Fuel pellet

diameter would be considered in

combination with initial enrichment

and cumulative burnup.

(b) Editorial corrections to the titles of

TS 3/4.9.14.1 and 3/4.9.14.2 would be

made for consistency with the TS

format.

(2) New TS 3/4.9.14.3 would be

added. The new TS would include:

(a) Requirements for acceptable fuel

storage in Region 1 of the SFP.

(b) An action statement, similar to

that for TS 3.9.14.1, requiring

suspension of all fuel movement and

crane operations except to move the

noncomplying fuel assemblies into an

acceptable pattern. The action statement

also requires verification of SFP boron

concentration at least once per 8 hours.

(c) A requirement, similar to that for

TS 4.9.14.1, for an evaluation that

considers enrichment, boron content,

and cumulative burnup of each fuel

assembly before storage in Region 1 of

the SFP.

(d) New Figure 3.9–3 for use in

determining the acceptability of storing

fuel in Region 1 of the SFP.

(3) The proposed changes to TS 5.3.1

are:

(a) The number of fuel rods in each

fuel assembly, nominal length of each

fuel rod, and maximum fuel enrichment

would be removed.

(b) The current allowance for fuel rod

substitutions as justified by analysis

would be clarified to specify that the

analysis be performed using NRC staffapproved methods.

(c) An allowance to use a limited

number of lead test assemblies in

nonlimiting core locations would be

added.

(d) The current specification requiring

Zircaloy-4 fuel cladding would be

changed to allow Zircaloy-4 or ZIRLO

cladding.

(4) The proposed changes to TS 5.6

are:

(a) TS 5.6.1.1 would be renumbered

TS 5.6.1 and the word ‘‘borated’’ would

be replaced with ‘‘unborated.’’

(b) A new requirement would be

added to specify the maximum fuel

enrichment allowed to be stored in the

fuel racks.

(c) TS 5.6.1.2 would be deleted.

(5) The associated Bases would also

be appropriately revised.

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. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

Analyses were performed to verify that an

increase in enrichment of the fuel from 4.5

weight percent U–235 to 5.0 weight percent

U–235 would not result in an inadvertent

criticality event in the new fuel storage racks

or the SFP. The analyses indicate that for the

new fuel racks, the keff will remain below

0.95 if flooded with non-borated water, and

below 0.98 if flooded with optimum-density

aqueous foam. The analyses indicate that for

the spent fuel racks, assuming credit for

soluble boron in accident scenarios, the keff

will remain below 0.95 as required.

The increase in the fuel enrichment from

4.5 weight percent U–235 to 5.0 weight

percent U–235 does not change any of the

external dimensional characteristics of the

fuel element, the fuel storage racks, or the

SFP itself. The accidents originally evaluated

considered those events that could lead to

fuel damage and release of radioactive

material primarily from mechanical means,

such as physical impact on the fuel or the

SFP. Because the physical design and

methods of operation are the same as

previously evaluated, there is no change in

the probability of occurrence of such events.

The maximum spent fuel gap activity and

the resulting offsite dose consequences after

a postulated fuel handling accident are

primarily dependent on fuel burnup, and are

not significantly affected by an increase in

fuel enrichment. For up to 5.0 weight percent

U–235 and 60,000 MWD/MTU burnup,

NUREG/CR–5009 indicates that fuel handling

accident offsite doses could increase by a

factor of 1.2, which indicates that doses

would still remain within 10 CFR Part 100

limits.

The Generic Letter 90–02 Supplement 1

change to TS 5.3.1 clarifies the requirements

associated with fuel reconstitution. It does

not change the methodology that would be

used to reconstitute fuel.

The use of ZIRLO cladding will not

increase the probability or consequences of

an accident, since it has improved

mechanical properties such as a lower

corrosion rate and reduced radiation-induced

growth.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

b. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The physical and mechanical parameters

associated with the fuel assemblies and spent

fuel racks are the same as previously

evaluated. Therefore, any malfunctions

related to the physical aspects of fuel storage

are the same as previously evaluated.

The conditions for fuel storage in the

proposed new TS 3.9.14.3 provide new

criteria for locations where a fuel assembly

could be incorrectly placed. However, the

incorrect placement of a fuel assembly has

been analyzed, and would not cause an

inadvertent criticality or any other accident.

The change to 5.0 weight percent U–235

does not result in physical alterations or

changes to the operation of the plant, or

change the method by which any safetyrelated system performs its function. The use

of ZIRLO cladding does not result in a

significant change to the plant.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

c. Does the change involve a significant

reduction in a margin of safety?

The acceptance criteria of a keff of 0.95 (or

0.98 for the new fuel rack optimum

moderation accident) provides the margin to

criticality. Analyses were performed that

conclude that the proposed changes to allow

up to 5.0 weight percent U–235 in the new

and spent fuel racks meet the acceptance

criteria. The use of ZIRLO cladding will not

reduce the protection of the public health or

safety, as indicated in the NRC’s revisions to

10 CFR 50.44 and 50.46 (57 FR 39355).

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 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: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps
Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices 11139

Department, San Luis Obispo, California

93407.

Attorney for licensee: Christopher J.

Warner, Esq., Pacific Gas and Electric

Company, P.O. Box 7442, San

Francisco, California 94120.

NRC Project Director: Theodore R.

Quay.

Pacific Gas and Electric Company,

Docket No. 50–133, Humboldt Bay

Power Plant, Unit 3, Humboldt County,

California

Date of amendment request:

November 23, 1994.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications Section VI,

‘‘Waste Disposal Systems,’’ regarding

radioactive effluent limitations and the

conditions for automatically pumping

the contents of the reactor caisson sump

to the outfall canal.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed revisions to the HBPP

Technical Specifications remove the

ambiguity in the guidelines for directing

caisson sump discharges to the outfall canal.

Additionally, the proposed revisions will

modify Section VI to be consistent with the

guidance provided by NRC Draft Generic

Letter for 10 CFR 20 Modification to

Technical Specifications (58 FR 68171, dated

December 23, 1993). These changes in

effluent limits are not related to the

probability or consequences of an accident.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed revisions to the HBPP

Technical Specifications are administrative

in nature and do not change the method by

which any safety-related system performs its

function.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does the change involve a significant

reduction in a margin of safety?

The proposed revisions to the HBPP

Technical Specifications do not affect the

margin of safety associated with parameters

for any accident analysis.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

analysis of the licensee and, based on

this review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Humboldt County Library, 636

F Street, Eureka, California 95501.

Attorney for licensee: Christopher J.

Warner, Esquire, Pacific Gas & Electric

Company, P.O. Box 7442, San

Francisco, California 94120.

NRC Project Director: Seymour H.

Weiss.

Pacific Gas and Electric Company,

Docket No. 50–133, Humboldt Bay

Power Plant, Unit 3, Humboldt County,

California

Date of amendment request:

November 23, 1994.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications Section

VII.C, Plant Staff, to decrease the

minimum staff requirements for the

shift operating organization from five to

two persons.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The probability or consequences of an

accident previously evaluated will not be

affected by the change in plant staffing. The

plant staff manning requirements for the shift

operating organization are being reduced to

reflect the condition of the plant in a

SAFSTOR mode. Previously evaluated

accidents do not require operator actions to

mitigate or reduce the consequences of

occurrence. Consequently, the change will

not affect the probability or consequences of

an accident occurring.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed revisions to the HBPP

Technical Specifications are administrative

in nature. Further, there would not be any

change in equipment or system function or

operation.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does the change involve a significant

reduction in a margin of safety?

The proposed revisions to the HBPP

Technical Specifications do not affect the

margin of safety of any accident analysis

since they do not affect the parameters for

any accident analysis, and they have no

effect on the current operating methodologies

or actions that govern plant performance.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

analysis of the licensee and, based on

this review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Humboldt County Library, 636

F Street, Eureka, California 95501.

Attorney for licensee: Christopher J.

Warner, Esquire, Pacific Gas & Electric

Company, P.O. Box 7442, San

Francisco, California 94120.

NRC Project Director: Seymour H.

Weiss.

PECO Energy Company, Public Service

Electric and Gas Company, Delmarva

Power and Light Company, and Atlantic

City Electric Company, Docket No. 50–

278, Peach Bottom Atomic Power

Station, Unit No. 3, York County,

Pennsylvania

Date of application for amendment:

January 13, 1995.

Description of amendment request:

The proposed changes revise Tables

3.7.1 and 3.7.4 to reflect a reduction in

the number of primary containment

power operated outboard valves for the

Traversing Incore Probe (TIP) probes,

and a redesignation of the containment

penetration numbers for the TIP ball,

shear, and check valves. The proposed

changes are a result of PBAPS

Modification P00068.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The TIP system does not serve as an

initiator or contributor to any accidents

previously evaluated. The system provides a

means of calibrating the Local Power Range

Monitors and supports thermal limit

calculations. The new system performs the

same function as the old one. It will provide

improved reliability and added redundancy

by allowing a complete flux mapping if a

detector or drive failure were to occur.

Installation of Modification P00068 and its

operation will not degrade any active or

passive equipment that responds to an

accident. These changes do not decrease the
11140 Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices

effectiveness of equipment relied upon to

mitigate the previously evaluated accidents.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any previously evaluated.

The modification is considered an

enhancement to the TIP system and does not

serve as an initiator or contributor to any of

the accidents previously evaluated. The

proposed changes do not introduce a new

mode of plant operation. The new system,

like the old one, is designed to keep the ball

valves closed upon reset of the Primary

Containment Isolation System (PCIS) logic.

The new TIP control console will respond to

a PCIS isolation signal in the same manner

as the old system.

Implementation of the proposed changes

will not affect the design function or

configuration of any component or introduce

any new operating scenarios or failure modes

or accident initiation.

Modification P00068 will not impair or

prevent safety systems from performing their

safety function. It will not make any changes

to the design function of the TIP system. The

classification of the TIP ball and shear valves

and their control circuitry will not change as

a result of this modification.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed changes do not involve a

significant reduction in a margin of safety.

The TIP system does not serve as an

initiator or contributor to any accidents

evaluated in the SAR [safety analysis report].

Modification P00068 is considered an

enhancement to the existing TIP system and

does not change its design function. The

reduction in the number of containment

penetrations from five to three does not

represent a reduction in a margin of safety

because of additional indexers in the new

system. The proposed changes do not

adversely affect the assumptions or sequence

of events used in any accident analysis.

Therefore, the proposed changes do not

involve a reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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 Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, Pennsylvania 17105.

Attorney for Licensee: J. W. Durham,

Sr., Esquire, Sr. V.P. and General

Counsel, PECO Energy Company, 2301

Market Street, Philadelphia,

Pennsylvania 19101.

NRC Project Director: John F. Stolz.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: June 13,

1994.

Description of amendment request:

The proposed change would remove

license condition 2.E from the Facility

Operating License. License Condition

2.E incorporated the requirements of

U.S. Department of Interior publication

‘‘Environmental Criteria for Electric

Transmission Systems’’—1970, which

applies to the construction cleanup,

restoration, and maintenance of

transmission lines.

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:

Operation of the FitzPatrick plant in

accordance with the proposed Amendment

would not involve a significant hazards

consideration as defined in 10 CFR 50.92,

since it would not:

(1) involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change will remove a license

condition unrelated to nuclear safety. License

condition 2.E incorporated into the Operating

License the requirements of U.S. Department

of Interior publication ‘‘Environmental

Criteria for Electric Transmission Systems’’—

1970. The goal of this standard is to

‘‘safeguard aesthetic and environmental

values within the constraints imposed by the

current state of high-voltage transmission

technology.’’ License condition 2.E addresses

the preservation of the environment and

natural resources. Removing this condition

from the Facility Operating License has no

bearing on plant safety or the health and

safety of the public considering its nonnuclear nature. The transmission line rightof-ways maintained by the [Power] Authority

[of the State of New York] are subject to

regulation by other State and Federal

Agencies. Removal of this license condition

will not affect operation of safety related

structures, systems or components nor affect

the quality assurance program at the

FitzPatrick plant. Therefore, the proposed

change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

(2) create the possibility of a new or

different kind of accident from any accident

previously evaluated.

License condition 2.E of the James A.

FitzPatrick Plant Operating License applies

to the construction cleanup, restoration, and

maintenance of transmission lines. The

Authority’s transmission lines are managed

under guidelines based on the ‘‘Generic

Transmission Line Right-of-Way

Management’’ plan requirements. The

requirements imposed by the plan on the

FitzPatrick transmission line right-of-ways

exceed those of the U.S. Department of

Interior publication referenced in license

condition 2.E in both scope and details.

Therefore, implementing the proposed

change will not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

(3) involve a significant reduction in a

margin of safety.

License condition 2.E of the James A.

FitzPatrick Plant Operating License applies

to the construction cleanup, restoration, and

maintenance of transmission lines. The

requirements imposed by this license

condition are unrelated to nuclear safety.

Continued operation of the plant without

Condition 2.E does not involve a significant

reduction in any margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mr. Charles M.

Pratt, 1633 Broadway, New York, New

York 10019.

NRC Project Director: Ledyard B.

Marsh.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request:

December 16, 1994; supplemented

February 10, 1995 (TS 94–07).

Description of amendment request:

The proposed change would reduce the

maximum allowed power levels and

more clearly specify the plant

conditions allowed by the technical

specifications for operation with one or

more main steam safety valves

inoperable. In addition, the Bases would

be revised to reflect these changes and

incorporate the revised methodology

used to establish the neutron flux

setpoints.

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:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of
Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices 11141

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

This change reduces the power level at

which the reactor may be operated with one

or more main steam safety valves (MSSVs)

inoperable, to ensure that the secondary

system is not overpressurized during the

most severe pressurization transient of the

secondary side. Additionally, this change

will combine the TS action statements for 3-

and 4-loop operation with one or more

MSSVs inoperable, revise the mode

requirements and times of Action Statement

3.7.1.1.a, and correct a reference in the bases

section to Table 3.7–1. Reduction of the high

neutron flux (HNF) trip setpoint will

continue to be used as the means to ensure

that the required reactor power level

reductions are met. Mode 3 will be limited

to application when the reactor trip breakers

(RTB) are closed. Lack of NIS trip setpoint

adjustments with the RTB open has no effect

on the accident analysis. There is no change

to the function of the MSSVs by the proposed

change. This change will not alter any

accident analysis assumptions or results for

SQN. The proposed change will reduce the

amount of relief capacity required to mitigate

the consequences of the transient by reducing

the total amount of energy in the primary

system. Therefore, this change will not

increase the probability of an accident.

This change is consistent with current SQN

accident analysis assumptions for the MSSVs

and does not change the containment

response for any design basis event.

Therefore, no change in the mitigation of an

accident will result from this proposed

change and no change will occur in the

consequences of any accident currently

analyzed.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

Inadvertent opening of a MSSV is currently

analyzed as an initiating event for accidental

depressurization of the main steam system.

The proposed change does not alter the

valves or any other plant component. The

valves will continue to perform as analyzed

in current accident analyses. The proposed

change will not create the possibility for any

new or different kind of accident.

By retaining the use of the HNF trip

setpoint reduction, no change is being

proposed in the methodology used to ensure

that power reductions are carried out;

therefore, this will not create the possibility

of placing the plant into any new unanalyzed

condition. Not adjusting the Nuclear

Instrumentation System trip setpoint with

the RTBs open will not create an accident.

The existing accident analysis is still

bounding.

Combining the separate action statements

for 3- and 4-loop operation into a single

action does not create the possibility for a

new or different kind of accident. Operation

with 4 loops will continue to be required in

Modes 1 and 2 by TS 3.4.1.1.

Operation with less than 4 loops will

continue to be governed by TS 3.4.1.2 in

Mode 3 and TS 3.4.1.3 in Mode 4. This

change will not place the plant in a

configuration not currently bounded by

existing accident analysis.

Revising the mode requirements and their

associated times, consistent with the

requirements in NUREG–1431, will continue

to ensure that if the unit is unable to comply

with the limiting condition for operation, the

unit will begin an orderly shutdown until a

mode is reached where the specification is

not applicable.

3. Involve a significant reduction in a

margin of safety.

The proposed change reduces the total

energy of the reactor coolant system that will

ensure the ability of the MSSVs to perform

their intended function as assumed in

current accident analyses. This change has

been evaluated on a generic basis for

Westinghouse Electric Corporation designed

4-loop nuclear steam supply systems. SQN

plant specific features have been evaluated

including power limit calculations and the

interaction of the reactor protection system

trip time delay and the anticipated transient

without scram mitigating system actuation

circuitry. Correcting this nonconservatism

restores the margin of safety to what was

originally envisioned. Therefore, the margin

of safety assumed in the accident analysis is

not reduced by this change.

Combining the separate action statements

for 3- and 4-loop operation into a single

action has no effect on the margin of safety

for 4-loop operation with one or more MSSVs

inoperable. Under the revised TS, 3-loop

operation with one or more MSSVs

inoperable would only be allowed in Mode

3, and 4-loop operation will be required in

Modes 1 and 2 in accordance with current

TSs 3.4.1.1 and 3.4.1.2.

Revising the mode requirements and their

associated times, consistent with the

requirements in NUREG–1431, will not

reduce the safety margin since the new

requirements will continue to place the unit

in a mode where the TS is no longer

applicable. The new completion times for

mode changes are reasonable, based on

operating experience, to reach the required

unit conditions from full power conditions in

an orderly manner without challenging unit

systems.

The margin of safety is unaffected by

modifying the limits of Mode 3 applicability

to require the RTBs to be closed as the

intended safety function is already

completed when they are open.

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request:

December 9, 1994, and January 27, 1995

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS)

Surveillance Requirement 4.6.1.2.a and

its associated Bases. The changes would

defer the next scheduled containment

integrated leak rate test (CILRT) for one

outage, from Refuel 7 (March 1995) to

Refuel 8 (scheduled for September

1996).

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 of occurrence or the

consequences of an accident or malfunction

of equipment important to safety previously

evaluated in the Safety Analysis Report.

The Callaway CILRT history provides

substantial justification for the proposed test

schedule. Three CILRTs have been performed

over a seven year period with successful

results. The tests indicate that Callaway has

a low leakage containment. There are no

structural mechanisms which would

adversely affect the structural capability of

the containment and that would be a factor

in extending the CILRT schedule by one

refueling outage.

A risk impact assessment was performed,

and a determination was made that there is

insignificant risk impact as a result of

changing the CILRT schedule. Containment

leak rate testing is not an initiator of any

accident, the proposed interval extension

does not affect reactor operations or the

accident analysis, and has no radiological

consequences. There will be no changes to 10

CFR 100 dose limits or the control room dose

limits. Extending the test interval will not, by

itself, increase the probability of a

malfunction of equipment important to

safety. Therefore, the proposed change will

not involve a significant 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

previously evaluated in the Safety Analysis

Report.

There are no design changes being made

that would create a new type of accident or

malfunction. The proposed change will not

alter the plant or the manner in which it is

operated. The change revises the schedule for

performing the periodic CILRT. The purpose

of the test is to provide periodic verification

of the leaktight integrity of the primary

reactor containment, and systems and

components which penetrate containment.

The tests assure that leakage through

containment and systems and components

penetrating containment will not exceed the

allowable leakage rate values associated with
11142 Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices

conditions resulting from an accident. The

change in schedule for performing the CILRT

will not adversely affect the containment

integrity in the event of an accident.

Therefore, the proposed change will not

create the possibility of a new or different

type of accident from any accident

previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The proposed change to the schedule for

performing the periodic Type A test does not

reduce the margin of safety assumed in the

accident analysis for any release of

radioactive materials or reduce any margin of

safety preserved by the technical

specifications. The methodology, acceptance

criteria, and the technical specification

leakage limits for the performance of the

Type A tests will not change. The Type A

tests will continue to be performed in

accordance with 10 CFR 50, Appendix J and

the Callaway Technical Specifications.

Therefore, the proposed change will not

involve a reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, N.W.,

Washington, DC 20037.

NRC Project Director: Leif J.

Norrholm.

Wisconsin Electric Power Company,

Docket Nos. 50–266 and 50–301, Point

Beach Nuclear Power Plant, Unit Nos. 1

and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of amendment request: January

24, 1995.

Description of amendment request:

The proposed amendment would

modify Technical Specification (TS)

Section 15.6.5, ‘‘Review and Audit,’’

and TS Section 15.7.8, ‘‘Administrative

Controls.’’ The quality assurance audit

frequencies would be removed, the

section on emergency plan reviews

would be removed, and the period for

radioactive effluent reporting would be

increased to annual. In addition, the

references to ‘‘Semiannual Monitoring

Report’’ would be changed to ‘‘Annual

Monitoring Report’’ throughout TS

Section 15.7.

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:

In accordance with the requirements of 10

CFR 50.91(a), Wisconsin Electric Power

Company (Licensee) has evaluated the

proposed changes against the standards of 10

CFR 50.92 and has determined that the

operation of Point Beach Nuclear Plant, Units

1 and 2, in accordance with the proposed

amendments, does not present a significant

hazards consideration.

A proposed facility operating license

amendment does not present a significant

hazards consideration if operation of the

facility in accordance with the proposed

amendment will not:

1. Create a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. Will not create a significant reduction in

a margin of safety.

The proposed changes are administrative

in nature. There is no physical change to the

facility, its systems, or its operation. Since

the changes will allow more flexibility in

assigning resources to work on poor or weak

performance areas, the plant safety will be

enhanced. Operation of PBNP in accordance

with the proposed amendments cannot create

an increase in the probability or

consequences of an accident previously

evaluated, create a new or different kind of

accident, or result in a significant reduction

in a margin of safety. Therefore, the proposed

changes do not present a significant hazards

consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Leif J.

Norrholm.

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.

Duke Power Company, et al., Docket No.

50–413, Catawba Nuclear Station, Unit

1, York County, South Carolina

Date of amendment request: October

18, 1994.

Description of amendment request:

The proposed amendment would

change Technical Specification 3.6.1.2

to defer the next scheduled containment

integrated leak rate test at Catawba Unit

1 for one outage, from the end-of-cycle

(EOC) 8 refueling outage (scheduled for

February 1995) to EOC 9 (scheduled for

June 1996).

Date of publication of individual

notice in Federal Register: February 6,

1995 (60 FR 7073).

Expiration date of individual notice:

March 8, 1995.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Duke Power Company, et al., Docket No.

50–413 Catawba Nuclear Station, Unit

1, York County, South Carolina

Date of amendment request:

November 29, 1994, as supplemented

January 12 and 27, 1995.

Description of amendment request:

The proposed amendment requested

renewal for Catawba Unit 1 Cycle 9

operation of the steam generator tube

inspection bobbin probe voltage-based

interim plugging criteria that had been

previously approved for Cycle 8.

Date of publication of individual

notice in Federal Register: February 9,

1995 (60 FR 7801).

Expiration date of individual notice:

March 13, 1995.

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Georgia Power Company, et al., Docket

Nos. 50–424 and 50–425, Vogtle Electric

Generating Plant, Units 1 and 2, Burke

County, Georgia,

Date of amendment request: January

20, 1995.

Description of amendment request:

The proposed amendment would revise

Technical Specification 6.4.1.2 to

provide a more accurate description of

the Plant Review Board composition.

Date of publication of individual

notice in Federal Register: February 6,

1995 (60 FR 7077).
Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices 11143

Expiration date of individual notice:

March 8, 1995.

Local Public Document Room

location: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50–498 and 50–499, South Texas

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request:

November 7, 1994, as supplemented by

letters dated December 20, 1994, and

January 23, 1995.

Brief description of amendment

request: The proposed amendments

would change the number of diesel

generators (emergency power supply)

required to be operable during Mode 6

with greater than or equal to 23 feet of

water above the reactor vessel flange,

from two to one. The amendments

would also allow limited substitution of

an alternate onsite emergency power

source for one of the two required diesel

generators, in Mode 5 and in Mode 6

with less than 23 feet of water. In

addition, changes to certain system

specifications that are affected by the

changes for the emergency power

supply were also proposed.

Date of individual notice in Federal

Register: January 30, 1995 (60 FR 5739).

Expiration date of individual notice:

March 1, 1995.

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, Texas

77488.

Illinois Power Company and Soyland

Power Cooperative, Inc., Docket No. 50–

461, Clinton Power Station, Unit No. 1,

DeWitt County, Illinois

Date of amendment request: January

27, 1995.

Brief description of amendment

request: The amendment modifies the

technical specifications (TSs) by

eliminating selected response time

testing as described in the BWROG

topical report NEDO–32291, ‘‘System

Analyses for Elimination of Selected

Response Time Testing Requirements.’’

The affected TSs are TS 3.3.1.1,

‘‘Reactor Protection System (RPS)

Instrumentation,’’ TS 3.3.5.1,

‘‘Emergency Core Cooling System

(ECCS) Instrumentation,’’ TS 3.3.6.1,

‘‘Primary Containment and Drywell

Isolation Instrumentation,’’ and TS

3.5.1, ‘‘ECCS—Operating.’’

Date of publication of individual

notice in Federal Register: February 3,

1995 (60 FR 6739).

Expiration date of individual notice:

March 6, 1995.

Local Public Document Room

location: Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, Illinois 61727.

Notice of Issuance of Amendments to

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

of 1954, as amended (the Act), and the

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

amendment, (2) the amendment, and (3)

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the

local public document rooms for the

particular facilities involved.

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Units 1, 2, and 3,

Maricopa County, Arizona

Date of application for amendments:

October 31, 1994, as supplemented by

letter dated December 28, 1994.

Brief description of amendments: The

amendments revise the refueling

machine overload cutoff limit from less

than or equal to 1556 pounds to less

than or equal to 1600 pounds. The

change was requested because design

and fabrication improvements have

increased the weight of the fuel

assembly.

Date of issuance: February 9, 1995.

Effective date: February 9, 1995, to be

implemented within 45 days of the date

of issuance.

Amendment Nos.: 89, 76, and 60.

Facility Operating License Nos. NPF–

41, NPF–51, and NPF–74: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 6, 1995 (60 FR 2160).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated February 9, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona

85004.

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Units 1, 2, and 3,

Maricopa County, Arizona

Date of application for amendments:

November 20, 1992, as supplemented by

letters dated October 22, 1993, and

November 30, 1994.

Brief description of amendments: The

amendments would increase the

allowable out-of-service time for the

core operating limit supervisory system

(COLSS) from 1 hour to 4 hours before

the more restrictive limits based on the

core protection calculators (CPCs) must

be applied.

Date of issuance: February 14, 1995.

Effective date: February 14, 1995.

Amendment Nos.: 90, 77, and 61.

Facility Operating License Nos. NPF–

41, NPF–51, and NPF–74: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 6, 1993 (58 FR 591)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated February 14, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona

85004.

Boston Edison Company, Docket No.

50–293, Pilgrim Nuclear Power Station,

Plymouth County, Massachusetts

Date of application for amendment:

September 6, 1994.
11144 Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices

Brief description of amendment: The

amendment would remove Technical

Specification Section 4.5.H.4 which

requires the testing and calibration of

pressure switches in certain emergency

core cooling system lines.

Date of issuance: February 2, 1995.

Effective date: February 2, 1995.

Amendment No.: 157.

Facility Operating License No. DPR–

35: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 26, 1994 (59 FR

53838). The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

October 26, 1994 (59 FR 53838).

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

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:

March 25, 1994, as supplemented on

July 29, 1994, and August 24, 1994.

Brief Description of amendments: The

amendments change the Technical

Specifications to correct several

typographical errors, to incorporate

material implicitly contained in a

footnote to an applicability statement, to

provide detailed labels for items listed

in a table, to correct the citation of

references, and to remove references to

the Rod Sequence Control System that

should have been included in a

previous change.

Date of issuance: February 1, 1995.

Effective date: February 1, 1995.

Amendment Nos.: 174 and 205.

Facility Operating License Nos. DPR–

71 and DPR–62. Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: May 25, 1994 (59 FR 27050).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated February 1, 1995.

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.

Carolina Power & Light Company,

Docket No. 50–261, H. B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of application for amendment:

December 12, 1994.

Brief description of amendment: The

amendment revises the containment

spray (CS) nozzle surveillance interval

from 5 to 10 years.

Date of issuance: February 10, 1995.

Effective date: February 10, 1995.

Amendment No.: 157.

Facility Operating License No. DPR–

23. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: January 4, 1995 (60 FR 497).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 10,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Hartsville Memorial Library,

147 West College, Hartsville, South

Carolina 29550.

Commonwealth Edison Company,

Docket Nos. 50–295 and 50–304, Zion

Nuclear Power Station, Units 1 and 2,

Lake County, Illinois

Date of application for amendments:

June 24, 1994.

Brief description of amendments: The

amendments revise the Technical

Specifications by deleting the

containment recirculation sump level

from Accident Monitoring

Instrumentation Tables 3.8.9–1 and

4.8.9–1.

Date of issuance: February 9, 1995.

Effective date: February 9, 1995.

Amendment Nos.: 160 and 148.

Facility Operating License Nos. DPR–

39 and DPR–48: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: July 20, 1994 (59 FR 37066).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated February 9,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois

60085.

Connecticut Yankee Atomic Power

Company, Docket No. 50–213, Haddam

Neck Plant, Middlesex County,

Connecticut, and Northeast Nuclear

Energy Company, Docket Nos. 50–245,

50–336, and 50–423, Millstone Nuclear

Power Station, Units 1, 2, and 3, New

London County, Connecticut

Date of application for amendments:

June 30, 1994, as supplemented

November 18, 1994, and January 12,

1995.

Brief description of amendments: The

amendments modify the Administrative

Controls Section of the Technical

Specifications by replacing the present

Nuclear Review Board (NRB) for the

Haddam Neck Plant, and the NRB and

site Nuclear Review Board for Millstone

Station with a Nuclear Safety

Assessment Board which will serve

Millstone Units 1, 2, and 3, and Haddam

Neck.

Date of issuance: February 14, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 181, 79, 184, 104.

Facility Operating License Nos. DPR–

61, DPR–21, DPR–65 AND NPF–49.

Amendments revised the Technical

Specifications.

The November 18, 1994, and January

12, 1995, submittals provided clarifying

information that did not change the

initial proposed no significant hazards

consideration determination.

Date of initial notice in Federal

Register: August 31, 1994 (59 FR

45021).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 14,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Russell Library, 123 Broad

Street, Middletown, CT 06457, for the

Haddam Neck Plant, and Learning

Resource Center, Three Rivers

Community-Technical College, Thames

Valley Campus, 574 New London

Turnpike, Norwich, CT 06360, for

Millstone 1, 2, and 3.

Duquesne Light Company, et al., Docket

No. 50–334, Beaver Valley Power

Station, Unit No. 1, Shippingport,

Pennsylvania

Date of application for amendment:

July 29, 1994, as supplemented in a

letter dated December 13, 1994.

Brief description of amendment: This

amendment revises Technical

Specifications (TSs) 3/4.4.5 and 3.4.6.2

including associated Bases 3/4.4.5 and

3/4.4.6.2 to allow the implementation of
Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices 11145

steam generator tube interim plugging

criteria (IPC) for the tube support plate

elevations during operating cycle 11.

The current TSs require that tubes with

imperfections exceeding 40 percent of

the nominal tube wall thickness be

removed from service. The IPC will

allow a test voltage-based criterion of

1.0 volts as determined by a bobbin

probe inspection of the tubes. Voltages

greater than 1.0 volt will be further

examined using a pancake coil probe.

Tubes showing flaw indications with a

bobbin voltage greater than 3.6 volts

will be plugged or repaired.

Date of issuance: February 3, 1995.

Effective date: February 3, 1995.

Amendment No: 184.

Facility Operating License No. DPR–

66. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 17, 1994 (59 FR

42337). The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

February 3, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 15001.

Entergy Operations, Inc., Docket No. 50–

313, Arkansas Nuclear One, Unit No. 1,

Pope County, Arkansas

Date of amendment request:

November 8, 1994.

Brief description of amendment: The

amendment revised the technical

specification section that describes the

frequency for performing the

containment integrated leak rate tests.

Date of issuance: February 6, 1995.

Effective date: February 6, 1995.

Amendment No.: 175.

Facility Operating License No. DPR–

51. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 4, 1995, (60 FR 502).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated February 6, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas

72801.

Florida Power and Light Company, et

al., Docket No. 50–389, St. Lucie Plant,

Unit No. 2, St. Lucie County, Florida

Date of application for amendment:

July 25, 1994.

Brief description of amendment: This

amendment will upgrade Technical

Specification 3/4.7.1.6 for the Main

Feedwater Line Isolation Valves to be

consistent with NUREG–1432, Standard

Technical Specifications for

Combustion Engineering Plants.

Date of Issuance: February 9, 1995.

Effective Date: February 9, 1995.

Amendment No.: 71.

Facility Operating License No. NPF–

16: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 31, 1994 (59 FR 45024)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated February 9, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954–9003.

Florida Power and Light Company, et

al., Docket Nos. 50–335 and 50–389, St.

Lucie Plant, Unit Nos. 1 and 2, St. Lucie

County, Florida

Date of application for amendments:

July 25, 1994.

Brief description of amendments:

These amendments implement GL 93–

05 Items 5.8, 6.1, 7.1 and 7.5.

Date of Issuance: February 9, 1995.

Effective Date: February 9, 1995.

Amendment Nos.: 133 and 72.

Facility Operating License Nos. DPR–

67 and NPF–16: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: August 31, 1994 (59 FR 45023)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated February 9, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954–9003.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50–498 and 50–499, South Texas

Project, Units 1 and 2, Matagorda

County, Texas.

Date of amendment request:

November 7, 1994.

Brief description of amendments: The

amendments permit both containment

personnel airlock doors to be open

while moving fuel during refueling

operations.

Date of issuance: February 2, 1995.

Effective date: February 2, 1995, to be

implemented within 30 days of

issuance.

Amendment Nos.: Unit 1—

Amendment No. 69; Unit 2—

Amendment No. 58.

Facility Operating License Nos. NPF–

76 and NPF–80. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 7, 1994 (59 FR

63123). The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

February 9, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, Texas

77488.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50–498 and 50–499, South Texas

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request:

November 8, 1994.

Brief description of amendments: The

amendments permit the substitution of

an extended range neutron flux monitor

for one of the source range neutron flux

monitors during refueling operations.

Date of issuance: February 13, 1995.

Effective date: February 13, 1995.

Amendment Nos.: Unit 1—

Amendment No. 70; Unit 2—

Amendment No. 59.

Facility Operating License Nos. NPF–

76 and NPF–80. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 7, 1994 (59 FR

63124). The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

February 13, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, Texas

77488.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50–498 and 50–499, South Texas

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: June 6,

1994, as supplemented by letters dated

November 17, 1994, and December 5,

1994.

Brief description of amendments: The

amendments modify Technical
11146 Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices

Specification 3/4.8.1.1, ‘‘A.C. Sources’’

by revising the action statements and

surveillance requirements for testing of

the standby diesel generators (SDGs).

The amendments eliminate excessive

and unnecessary testing of the SDGs

consistent with the guidance provided

in NUREG–1366, ‘‘Improvements to

Technical Specifications Surveillance

Requirements,’’ NUREG–1431,

‘‘Standard Technical Specifications for

Westinghouse Plants,’’ Generic Letter

84–15, ‘‘Proposed Staff Actions to

Improve and Maintain Diesel Generator

Reliability,’’ and Generic Letter 93–05,

‘‘Line-Item Technical Specifications

Improvements to Reduce Surveillance

Requirements for Testing During Power

Operation.’’ The changes include: (1)

eliminating the requirement to

demonstrate the operability of an

operable SDG whenever an offsite AC

power source is determined to be

inoperable, or whenever a support

system or an independently testable

component of another SDG is

inoperable, (2) eliminating the

requirement to load the diesel in 10

minutes during testing, (3) replacing the

minimum required loading for testing

with a load band, (4) relocating some

surveillance requirements to the Diesel

Fuel Oil Testing Program, and (5)

eliminating unnecessary loss-of-offsite

power tests.

Date of issuance: February 2, 1995.

Effective date: February 2, 1995, to be

implemented within 60 days of

issuance.

Amendment Nos.: Unit 1—

Amendment No. 68; Unit 2—

Amendment No. 57.

Facility Operating License Nos. NPF–

76 and NPF–80. Amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 20, 1994 (59 FR 37073).

The November 17, 1994, and December

5, 1994, submittals provided clarifying

information and did not change the

original no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated February 2,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

Location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, Texas

77488.

Indiana Michigan Power Company,

Docket Nos. 50–315 and 50–316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of application for amendments:

August 3, 1994.

Brief description of amendments: The

amendments relocate the Radiological

Effluent Technical Specifications to

other controlled documents consistent

with NRC Generic Letter 89–01.

Date of issuance: February 10, 1995.

Effective date: February 10, 1995.

Amendment Nos.: 189 and 175.

Facility Operating License Nos. DPR–

58 and DPR–74. Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55873).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated February 10,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085.

Maine Yankee Atomic Power Company,

Docket No. 50–309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of application for amendment:

October 24, 1994, as supplemented by

letter dated December 16, 1994.

Brief description of amendment: This

amendment modifies Technical

Specifications Table 4.1–3 surveillance

requirements for the new emergency

feedwater flow instrumentation.

Specifically, the currently installed

analog feedwater flow transmitters are

to be replaced by new, digital-type flow

transmitters. The new digital flow

emergency feedwater flow transmitters

are continuously self-checking and have

a recommended calibration interval of 9

years. The licensee will verify flow

whenever the system operates and send

one transmitter back to the

manufacturer for recalibration every

refueling outage.

Date of issuance: February 15, 1995.

Effective date: February 15, 1995.

Amendment No.: 147.

Facility Operating License No. DPR–

36: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 7, 1994 (59 FR

63124). The December 16, 1994, letter

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 15,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, Maine

04578.

Maine Yankee Atomic Power Company,

Docket No. 50–309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of application for amendment:

May 25, 1994, as supplemented

September 1, 1994, and January 13,

1995.

Brief description of amendment: This

amendment allows (1) entry through an

operable personnel air lock hatch to

perform surveillance testing, repair an

inoperable hatch, or perform other

necessary activities inside containment;

(2) update plant Technical

Specifications to reflect a previous

change to the list of containment

boundary valves; (3) add a new

exception to allow quarterly

surveillance testing of the excess flow

check valves; (4) add a new exception

to allow periodic preventive

maintenance on control room

ventilation lasting up to 30 minutes per

calendar quarter, without a written

report of such inoperability; and (5)

make related administrative changes to

reflect and clarify items 1 through 4

above.

Date of issuance: February 10, 1995.

Effective date: February 10, 1995.

Amendment No.: 146.

Facility Operating License No. DPR–

36: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 22, 1994 (59 FR 32231).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated February 10, 1995.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, Maine

04578.

Niagara Mohawk Power Corporation,

Docket No. 50–410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County,

New York

Date of application for amendment:

November 3, 1993.

Brief description of amendment: The

amendment revises License Condition

2.C.(4), ‘‘Turbine System Maintenance

Program,’’ and deletes Technical

Specification (TS) 3/4.3.8, ‘‘Turbine
Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices 11147

Overspeed Protection System,’’ and its

associated Bases. The revision to

License Condition 2.C.(4) indicates that

the requirements of this license

condition have been satisfied. The

deletion of TS 3/4.3.8 and its associated

Bases provides Niagara Mohawk Power

Corporation the flexibility to implement

the manufacturer’s recommendations for

turbine steam valve surveillance test

requirements. These test requirements

will be contained in the Updated Safety

Analysis Report.

Date of issuance: February 14, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 63.

Facility Operating License No. NPF–

69: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: December 8, 1993 (58 FR

64611). The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

February 14, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Northeast Nuclear Energy Company, et

al., Docket No. 50–336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of application for amendment:

April 25, 1994.

Brief description of amendment: The

amendment changes the Technical

Specifications concerning four related

issues: (1) power-operated relief valve

and block valve reliability; (2) lowtemperature overpressure protection; (3)

boron dilution; and (4) shutdown risk

management.

Date of issuance: February 15, 1995.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 185.

Facility Operating License No. DPR–

65. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 25, 1994 (59 FR 27060).

The September 21, 1994, letter provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 15,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Thames Valley State Technical College,

574 New London Turnpike, Norwich,

Connecticut 06360.

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: June 18,

1993, as supplemented by letter dated

November 23, 1994.

Description of amendment request:

The amendment revises the Appendix A

Technical Specifications (TS) relating to

the Independent Safety Engineering

Group. Specifically, the amendment

revises the title of TS 6.2.3 from

Independent Safety Engineering Group

to Independent Technical Reviews, and

replaces the requirements for the five

person Independent Safety Engineering

Group with requirements relating to a

technical review program to perform

independent technical reviews.

Date of issuance: February 14, 1995.

Effective date: As of the date of

issuance to be implemented within 60

days.

Amendment No.: 35.

Facility Operating License No. NPF–

86. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 18, 1993 (58 FR

43927). The licensee’s letter dated

November 23, 1994, provided a minor

revision to the application but does 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 February 14,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Exeter Public Library, 47 Front

Street, Exeter, NH 03833.

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 application for amendments:

October 21, 1994.

Brief description of amendments:

These amendments add a test exception

to allow reactor coolant temperatures up

to 212 degrees F during hydrostatic or

inservice leak testing while in

OPERATIONAL CONDITION 4 without

entering OPERATIONAL CONDITION 3.

Date of issuance: February 13, 1995.

Effective date: To be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 142 and 112.

Facility Operating License Nos. NPF–

14 and NPF–22. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 22, 1994 (59 FR

66057). The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

February 13, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre,

Pennsylvania 18701.

Philadelphia Electric Company, Docket

No. 50–353, Limerick Generating

Station, Unit 2, Montgomery County,

Pennsylvania

Date of application for amendment:

December 9, 1993, as supplemented by

letters dated July 5, September 9,

October 19, November 15, and

December 2, 1994, January 6 and

January 23, 1995. The supplemental

letters provided clarifying information

that did not change the initial no

significant hazards consideration

determination.

Brief description of amendment: This

amendment raises the authorized

maximum power level from 3293 MWt

to a new limit of 3458 MWt. The

amendment also approves changes to

the Technical Specifications to

implement uprated power operation.

Date of issuance: February 16, 1995.

Effective date: This license

amendment is effective as of its date of

issuance and is to be implemented prior

to startup in Cycle 4.

Amendment No.: 51.

Facility Operating License No. NPF–

85. This amendment revised the

Technical Specifications and License.

Date of initial notice in Federal

Register: February 16, 1994 (59 FR

7695). The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

February 16, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Philadelphia Electric Company, Docket

Nos. 50–352 and 50–353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of application for amendments:

October 29, 1993.

Brief description of amendments:

These amendments eliminate the main
11148 Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices

steamline radiation monitoring system

high radiation trip function for initiating

(1) an automatic reactor scram and

automatic closure of the main steamline

isolation valves, and (2) automatic

closure of the main steamline drain

valves, main steam and reactor water

sample line valves. The amendments

also approve the relocation of portions

of the information contained in the

Bases section.

Date of issuance: February 16, 1995.

Effective date: February 16, 1995.

Amendment Nos. 89 and 52.

Facility Operating License Nos. NPF–

39 and NPF–85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: January 5, 1994 (59 FR 624).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated February 16, 1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Rochester Gas and Electric Corporation,

Docket No. 50–244, R. E. Ginna Nuclear

Power Plant, Wayne County, New York

Date of application for amendment:

May 13, 1994, as supplemented June 24

and September 27, 1994.

Brief description of amendment: The

amendment proposes to amend

Appendix A of Operating License DPR–

18 to revise Section 6.0 ‘‘Administrative

Controls’’ of the Ginna Technical

Specifications (TSs) and would change

the title of Senior Vice President,

Production and Engineering, include a

provision to allow future title changes

without license amendment, and

implement those changes in NUREG–

1431 ‘‘Standard Technical

Specification—Westinghouse Plants,’’

dated September 1992, by relocating to

licensee controlled documents those

specifications controlled by regulations

and the existing review and audit

requirements. The remainder of this

amendment request will be reviewed at

a later date.

Date of issuance: February 6, 1995.

Effective date: February 6, 1995.

Amendment No.: 58.

Facility Operating License No. DPR–

18: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 20, 1994 (59 FR 37084).

The June 24, 1994, submittal provided

information which did not change the

initial no significant hazards

consideration determination. The

licensee’s submittal of September 27,

1994, limited, but did not change, the

licensee’s previously requested TS

changes of May 13, 1994.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 6,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Rochester Public Library, 115

South Avenue, Rochester, New York

14610.

Southern California Edison Company, et

al., Docket Nos. 50–361 and 50–362,

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

California

Date of application for amendments:

December 30, 1993, as supplemented by

letters dated June 3, 1994, August 25,

1994, and January 3, 19, and 30, 1995.

Brief description of amendments:

These amendments will revise TS Table

3.3–1, ‘‘Reactor Protective

Instrumentation,’’ to allow the use of the

source range neutron flux monitors in

place of safety related excore monitors

in Modes 3, 4, and 5, with the reactor

trip circuit breakers open or the Control

Element Assembly (CEA) Drive System

not capable of CEA withdrawal, for the

purpose of monitoring core reactive

changes.

Date of issuance: February 13, 1995.

Effective date: February 13, 1995.

Amendment Nos.: 115 and 104.

Facility Operating License Nos. NPF–

10 and NPF–15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: September 28, 1994 (59 FR

49434). The additional information

contained in the January 3, 19, and 30,

1995, letters were clarifying in nature,

within the scope of the initial notice

and did not affect the NRC staff’s

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated February 13,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Main Library, University of

California, P.O. Box 19557, Irvine,

California 92713.

Southern California Edison Company, et

al., Docket Nos. 50–361 and 50–362,

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

California

Date of application for amendments:

July 28, 1994, as supplemented by

letters dated January 30 and February

13, 1995.

Brief description of amendments:

These amendments propose to revise

Technical Specification (TS) 3.9.8.1

‘‘Shutdown Cooling and Coolant

Circulation—High Water Level,’’ TS

3.9.8.2 ‘‘Shutdown Cooling and Coolant

Circulation—Low Water Level,’’ and

their Bases to facilitate testing of lowpressure safety injection system

components and permit additional

flexibility in scheduling maintenance on

the shutdown cooling system.

Date of issuance: February 15, 1995.

Effective date: February 15, 1995.

Amendment Nos.: 116 and 105.

Facility Operating License Nos. NPF–

10 and NPF–15: The amendments

revised the Technical Specifications on

a one-time basis for each unit.

Date of initial notice in Federal

Register: October 12, 1994 (59 FR

51627). The additional information

contained in the supplemental letters

dated January 30 and February 13, 1995,

served to clarify the amendments, was

within the scope of the initial notice,

and did not affect the Commission’s

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated February 15,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Main Library, University of

California, P. O. Box 19557, Irvine,

California 92713.

South Carolina Electric & Gas Company,

South Carolina Public Service

Authority, Docket No. 50–395, Virgil C.

Summer Nuclear Station, Unit No. 1,

Fairfield County, South Carolina

Date of application for amendment:

October 17, 1994, as supplemented

January 30, 1995.

Brief description of amendment: The

amendment changes the Technical

Specifications to relocate the seismic

monitoring instrumentation (SMI)

Limiting Condition for Operation (LCO),

Surveillance Requirements (SRs), and

associated tables and bases contained in

Technical Specifications (TS) sections

3.3.3.3 and 4.3.3.3 to the Final Safety

Analysis Report (FSAR) or an equivalent

controlled document.

Date of issuance: February 15, 1995.

Effective date: February 15, 1995.

Amendment No.: 122.

Facility Operating License No. NPF–

12. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: November 8, 1994 (59 FR

55717). The January 30, 1995,

supplement did not affect the staff’s
Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices 11149

finding of no significant hazards

consideration.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 15,

1995.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Fairfield County Library, 300

Washington Street, Winnsboro, SC

29180.

Tennessee Valley Authority, Docket

Nos. 50–259, 50–260, and 50–296,

Browns Ferry Nuclear Plant, Units 1, 2,

and 3, Limestone County, Alabama

Date of application for amendments:

September 29, 1993.

Brief Description of amendment: The

proposed changes revise standards for

testing of charcoal used for removal of

radioactive iodine in ventilation

systems at the Browns Ferry Nuclear

Plant.

Date of issuance: February 13, 1995.

Effective Date: February 13, 1995.

Amendment Nos.: 215, 231 and 188.

Facility Operating License Nos. DPR–

33, DPR–52 and DPR–68: Amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 22, 1993 (58 FR

67862). The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

February 13, 1995.

No significant hazards consideration

comments received: None.

Local Public Document Room

Location: Athens Public library, South

Street, Athens, Alabama 35611.

Tennessee Valley Authority, Docket No.

50–296, Browns Ferry Nuclear Plant,

Unit 3, Limestone County, Alabama

Date of application for amendment:

March 29, 1994.

Brief Description of amendment: The

amendment adds requirements for load

shedding components being added to

ensure that emergency diesel generators

are not overloaded during design basis

accidents.

Date of issuance: February 14, 1995.

Effective Date: February 14, 1995.

Amendment No.: 189.

Facility Operating License No. DPR–

68: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 3, 1994 (59 FR 39597).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated February 14, 1995.

No significant hazards consideration

comments received: None.

Local Public Document Room

Location: Athens Public library, South

Street, Athens, Alabama 35611.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

December 16, 1994; supplemented

January 19, 1995 (TS 94–16).

Brief description of amendments: The

amendments remove the 900 rpm

emergency diesel generator surveillance

test criteria and a requirement that the

plant be shutdown for performance of

the interdependence diesel generator

tests.

Date of issuance: February 9, 1995.

Effective date: February 9, 1995.

Amendment Nos.: 195 and 186.

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revise the

technical specifications.

Date of initial notice in Federal

Register: December 29, 1994 (59 FR

67350). The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

February 9, 1995.

No significant hazards consideration

comments received: None.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402.

Notice of Issuance of Amendments to

Facility Operating Licenses and Final

Determination of No Significant

Hazards Consideration and

Opportunity for a Hearing (Exigent

Public Announcement or Emergency

Circumstances)

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 for the

amendment complies with the

standards and requirements of the

Atomic Energy Act of 1954, as amended

(the Act), and the Commission’s rules

and regulations. The Commission has

made appropriate findings as required

by the Act and the Commission’s rules

and regulations in 10 CFR Chapter I,

which are set forth in the license

amendment.

Because of exigent or emergency

circumstances associated with the date

the amendment was needed, there was

not time for the Commission to publish,

for public comment before issuance, its

usual 30-day Notice of Consideration of

Issuance of Amendment, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for a

Hearing.

For exigent circumstances, the

Commission has either issued a Federal

Register notice providing opportunity

for public comment or has used local

media to provide notice to the public in

the area surrounding a licensee’s facility

of the licensee’s application and of the

Commission’s proposed determination

of no significant hazards consideration.

The Commission has provided a

reasonable opportunity for the public to

comment, using its best efforts to make

available to the public means of

communication for the public to

respond quickly, and in the case of

telephone comments, the comments

have been recorded or transcribed as

appropriate and the licensee has been

informed of the public comments.

In circumstances where failure to act

in a timely way would have resulted, for

example, in derating or shutdown of a

nuclear power plant or in prevention of

either resumption of operation or of

increase in power output up to the

plant’s licensed power level, the

Commission may not have had an

opportunity to provide for public

comment on its no significant hazards

consideration determination. In such

case, the license amendment has been

issued without opportunity for

comment. If there has been some time

for public comment but less than 30

days, the Commission may provide an

opportunity for public comment. If

comments have been requested, it is so

stated. In either event, the State has

been consulted by telephone whenever

possible.

Under its regulations, the Commission

may issue and make an amendment

immediately effective, notwithstanding

the pendency before it of a request for

a hearing from any person, in advance

of the holding and completion of any

required hearing, where it has

determined that no significant hazards

consideration is involved.

The Commission has applied the

standards of 10 CFR 50.92 and has made

a final determination that the

amendment involves no significant

hazards consideration. The basis for this

determination is contained in the

documents related to this action.

Accordingly, the amendments have

been issued and made effective 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
11150 Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the application for

amendment, (2) the amendment to

Facility Operating License, 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 room for the

particular facility involved.

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By

March 31, 1995, the licensee may file a

request for a hearing with respect to

issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

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, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

Union at 1–(800) 248–5100 (in Missouri

1–(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, 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 the

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

Commonwealth Edison Company,

Docket No. 50–374, LaSalle County

Station, Unit 2, LaSalle County, Illinois

Date of application for amendment:

January 30, 1995.

Brief description of amendment: The

amendment adds a footnote to

Technical Specification Table 4.3.1.1–1

to allow a one-time extension of the

surveillance interval for the main steam

line isolation valve (MSIV) closure

reactor protection system channel

functional test. This extension averts the

need to perform the functional test prior

to the start of the upcoming Unit 2

refueling outage.

Date of Issuance: February 15, 1995.

Effective date: Immediately and shall

be implemented prior to 2:45 a.m. CST

on February 15, 1995.

Amendment No.: 86.

Facility Operating License No. NPF–

18: The amendment revised the

Technical Specifications.

Press release issued requesting

comments as to proposed no siginificant

hazards consideration: Yes. February 6,

1995, Morris Daily Herald; Ottawa Daily

Times; and Streator Times-Press.

Comments received: No. The

Commission’s related evaluation of the

amendment, finding of exigent

circumstances, consultation with the

State of Illinois and final determination

of no significant hazards consideration
Federal Register / Vol. 60, No. 40 / Wednesday, March 1, 1995 / Notices 11151

are contained in a Safety Evaluation

dated February 14, 1995.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60690.

Local Public Document Room

location: Public Library of Illinois

Valley Community College, Rural Route

No. 1, Oglesby, Illinois 61348.

NRC Project Director: Robert A. Capra.

Dated at Rockville, Maryland, this 21st day

of February 1995.

For the Nuclear Regulatory Commission.

John N. Hannon,

Acting Director, Division of Reactor Projects—

III/IV, Office of Nuclear Reactor Regulation.

[FR Doc. 95–4870 Filed 2–28–95; 8:45 am]

BILLING CODE 7590–01–P

[Docket Nos. 50–245, 50–336, 50–423]

Northeast Utilities; Issuance of

Director’s Decision Under 10 CFR

2.206

[Millstone Nuclear Power Station]

[License Nos. DPR–21, DPR–65, NPF–49]

Notice is hereby given that the Acting

Director, Office of Enforcement, has

issued a decision concerning the

Petitions filed by Ms. Carmela V. Marien

and Ms. Marianne W. Nericcio on

August 21, 1993. The Petitions

requested that the NRC initiate an

investigation and accelerated

enforcement action against Northeast

Utilities (Licensee) for willful violation

of the employee protection provisions of

10 CFR 50.7.

After due consideration of Petitioner’s

assertions, the Acting Director, Office of

Enforcement, has denied the Petitions.

The reasons for the denial are explained

in the ‘‘Director’s Decision under 10

CFR 2.206’’ (DD–95–04) which is

available for public inspection in the

Commission’s Public Document Room

at 2120 L Street NW., Washington, DC

20555.

A copy of this decision will be filed

with the Secretary for the Commission’s

review in accordance with 10 CFR

2.206. As provided by this regulation,

the decision will constitute the final

action of the Commission 25 days after

the date of issuance of the decision

unless the Commission on its own

motion institutes a review of the

decision within that time.

Dated at Rockville, Maryland this 22nd day

of February 1994.

For the Nuclear Regulatory Commission.

Joseph R. Gray,

Acting Director, Office of Enforcement.

[FR Doc. 95–4978 Filed 2–28–95; 8:45 am]

BILLING CODE 7590–01–M

[Docket No. 50–219]

GPU Nuclear Corporation; Notice of

Issuance of Amendment to Facility

Operating License

The U.S. Nuclear Regulatory

Commission (Commission) has issued

Amendment No. 177 to Facility

Operating License No. NPF–16 issued to

GPU Nuclear Corporation (the licensee),

which revised the Technical

Specifications for operation of the

Oyster Creek Nuclear Generating Station

located in Ocean County, New Jersey.

The amendment is effective as of the

date of issuance, to be implemented

within 60 days of issuance.

The amendment revises Technical

Specification 2.3.D to change the

setpoints ‘‘Reactor High Pressure, Relief

Valve Initiation’’ by increasing the

setpoint value by 15 psig for each of the

Electromatic Relief Valves in the

Automatic Depressurization System.

The application for the amendment

complies with the standards and

requirements of the Atomic Energy Act

of 1954, as amended (the Act), and the

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment and Opportunity for

Hearing in connection with this action

was published in the Federal Register

on July 5, 1994 (59 FR 34453). No

request for a hearing or petition for

leave to intervene was filed following

this notice.

The Commission has prepared an

Environmental Assessment related to

the action and has determined not to

prepare an environmental impact

statement. Based upon the

environmental assessment, the

Commission has concluded that the

issuance of the amendment will not

have a significant effect on the quality

of the human environment (60 FR 9056).

For further details with respects to the

action see (1) The application for

amendment dated June 15, 1994, as

supplemented September 23, and

November 23, 1994, (2) Amendment No.

177 to License No. DPR–16, (3) the

Commission’s related Safety Evaluation,

and (4) the Commission’s

Environmental Assessment. 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 room located at

the Ocean County Library, Reference

Department, 101 Washington Street,

Toms River, NJ 08753.

Dated at Rockville, Maryland, this 21st day

of February 1995.

For the Nuclear Regulatory Commission.

Alexander W. Dromerick, Sr.

Project Manager, Project Directorate I–4,

Division of Reactor Projects—I/II, Office of

Nuclear Reactor Regulation.

[FR Doc. 95–4977 Filed 2–28–95; 8:45 am]

BILLING CODE 7590–01–M

[Docket No. 50–483]

Union Electric Company;

Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for Hearing

The U.S. Nuclear Regulatory

Commission (the Commission) is

considering issuance of an amendment

to Facility Operating License No. NPF–

30, issued to Union Electric Company

(the licensee), for operation of the

Callaway plant, located in Callaway

County, Missouri.

The proposed amendment would

modify Technical Specification (TS)

Section 3/4.9.1 to establish

administrative controls to address a

possible boron dilution event directly

from the reactor makeup water (RMW)

system. An unreviewed safety question

was involved with the use of RMW to

rinse items removed from the refueling

pool and to spray down the refueling

pool walls during the pool drain

evolution. The use of RMW in prior

refueling outages during these Mode 6

activities raised the possibility of a

different type of accident than any

previously evaluated in the Callaway

Final Safety Analysis Report (FSAR).

FSAR Section 15.4.6.2 currently states

that administrative controls during

Mode 6, i.e., closing and locking

dilution source manual valves, preclude

an inadvertent dilution of the boron

concentration of the primary system.

Since these valve closures do not

preclude the potential dilution scenario

described above, different procedural

controls are required to ensure that LCO

3.9.1 boron concentration limit of 2000

ppm is met.

NRC Generic Letter 85–05,

‘‘Inadvertent Boron Dilution Events,’’

January 1985, and NSAC–183, ‘‘Risk of

PWR Inadvertent Criticality During

Shutdown and Refueling,’’ dated

December 1992, documents the

technical justification for determining

that boron dilution events are selflimiting. Based on the analyses

provided in these documents, the staff’s

acceptance criteria remains valid for the

different boron dilution transient (i.e.,

that gradual boron dilution events are