25696 Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices

Dated at Rockville, Maryland this 15th day

of May 1996.

James Lieberman,

Director, Office of Enforcement.

[FR Doc. 96–12810 Filed 5–21–96; 8:45 am]

BILLING CODE 7590–01–P

UNITED STATES NUCLEAR

REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to

Facility Operating Licenses Involving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97-415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from April 27,

1996, through May 10, 1996. The last

biweekly notice was published on May

8, 1996 (61 FR 20842).

Notice Of Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

Rockville, Maryland from 7:30 a.m. to

4:15 p.m. Federal workdays. Copies of

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By June 21, 1996, 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
Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices 25697

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, Attention:

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 20555-0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Baltimore Gas and Electric Company,

Docket Nos. 50-317 and 50-318, Calvert

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of amendments request: April 5,

1996

Description of amendments request:

Pursuant to 10 CFR 50.80 and 50.90, the

Baltimore Gas and Electric Company

(BGE) hereby requests the transfer and

amendment of Operating License Nos.

DPR-53 and DPR-69 for Calvert Cliffs

Nuclear Power Plant, Unit Nos. 1 and 2.

The proposed license transfers and

amendments are requested as part of the

pending merger between BGE and

Potomac Electric Power Company into

Constellation Energy Corporation. The

proposed license transfers would

transfer authority to possess and operate

Calvert Cliffs from BGE to Constellation

Energy Corporation. The proposed

amendments would change the licenses

as well as the related Technical

Specifications, to reflect this transfer by

submitting Constellation Energy

Corporation in place of BGE as the

licensee for Calvert Cliffs.

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. Would not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The proposed amendment will change the

name of the licensee authorized to possess

and operate Calvert Cliffs Nuclear Power

Plant from Baltimore Gas and Electric

Company (BGE) to Constellation Energy

Corporation. This amendment request is

necessary because of a proposed merger of

BGE and Potomac Electric Power Company

into Constellation Energy Corporation. As a

result of the savings achieved through a

reduction in operating costs due to the

merger, Constellation Energy Corporation

will have the financial resources to possess

and operate Calvert Cliffs.

In addition, Constellation Energy

Corporation personnel will be technically

qualified to operate the plant. Baltimore Gas

and Electric Company nuclear personnel

have been named to management positions in

Constellation Energy Corporation, and will

remain responsible for Calvert Cliffs

operation and maintenance. The proposed

amendment involves no changes in the

training program or operating organization

for Calvert Cliffs.

The proposed amendment does not require

any physical change to the facilities or

substantive modifications to the Technical

Specifications or to procedures. The

proposed change does not increase the

probability of an accident previously

evaluated because it does not affect any

initiators in any previously evaluated

accidents. The proposed change does not

increase the consequences of an accident

previously evaluated because it does not

affect any of the items on which the

consequences depend.

Therefore, the proposed amendment does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Would not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

The proposed amendment does not modify

the plant’s configuration or operations. As a

result, no new accident initiators are

introduced. Therefore, the proposed

amendment does not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

3. Would not involve a significant

reduction in a margin of safety.

This amendment request is necessary

because of a proposed merger of BGE and

Potomac Electric Power Company into

Constellation Energy Corporation. As a result

of the savings achieved through a reduction

in operating costs due to the merger,

Constellation Energy Corporation will have

the financial resources to possess and operate

Calvert Cliffs. Also, Constellation Energy

Corporation personnel will be technically

qualified to operate the plant. Baltimore Gas

and Electric Company nuclear personnel

have been named to management positions in

Constellation Energy Corporation, and will

remain responsible for Calvert Cliffs’

operation and maintenance. The proposed

amendment involves no changes in the

training program or operating organization

for Calvert Cliffs. In addition, the proposed

amendment to substitute Constellation

Energy Corporation for BGE does not result

in any changes to the physical design or

operation of the plant. Therefore, 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 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendments request

involves no significant hazards

consideration.

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.
25698 Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices

Attorney for licensee: Jay E. Silbert,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Susan Frant

Shankman, Acting

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 amendments request: April 2,

1996

Description of amendments request:

The proposed amendments revise the

Brunswick Steam Electric Plant, Units 1

and 2, Technical Specifications (TS) to

allow uprate of the units to 105 percent

of rated thermal power.

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:

. May the proposed activity involve a

significant increase in the probability or

consequences of an accident evaluated

previously in the Safety Analysis Report?

The increase in power level, steam flow,

feedwater flow and associated instrument

setpoint changes will not significantly

increase the probability or consequences of

an accident previously evaluated.

The probability (frequency of occurrence)

of Design Basis Accidents occurring is not

affected by the increase in power level, as

plant equipment will remain in compliance

with the applicable regulatory criteria (ASME

Codes, IEEE Standards, NEMA Standards,

Regulatory Guide criteria, etc.). The physical

plant changes necessary to support power

uprate include instrument setpoint changes,

indicating meter scale changes for the RWCU

[reactor water cleanup] System flow and

Main Steam Flow indicators, Leak Detection,

Process Computer, ERFIS [emergency

response facility information system], and

Feedwater System software changes, and

SRV [safety/relief valve] setpoint changes.

The setpoints were calculated in accordance

with the CP&L Setpoint Methodology.

Utilizing this methodology ensures scram

setpoints (instrument settings that initiate

automatic plant shutdowns) will be

established such that there is no significant

increase in scram frequency due to uprate.

No new challenges to safety related

equipment will result from power uprate.

The changes in consequences of

hypothetical accidents which would occur

from 102% of the uprated power (2609 MWt),

compared to those previously evaluated from

[greater than or equal to] 102% of the original

power (2485 MWt), are not significant,

because the accident evaluations at uprated

power will not result in exceeding the NRC

approved acceptance limits. The spectrum of

hypothetical accidents and transients has

been investigated, and those accidents/

transients currently evaluated in the UFSAR

[Updated Final Safety Analysis Report] were

shown to meet the plant’s current regulatory

criteria at uprated conditions (105%). In the

area of core design, for example, the fuel

operating limits will still be met at the

uprated power level, and fuel reload analyses

show plant transients will still meet the

criteria accepted by the NRC as specified in

NEDO-24011, ‘‘GESTAR II.’’ Challenges to

fuel or ECCS [emergency core cooling

system] performance have been evaluated

and shown to meet the criteria of 10CFR50

Appendix K. Challenges to the containment

have been evaluated and still meet 10CFR50

Appendix A Criterion 38, Long Term

Cooling, and Criterion 50, Containment.

Bounding events involving radiological

releases have been evaluated and were

shown to be well within the criteria of

10CFR100.

2. May the proposed activity create the

possibility of a new or different kind of

accident from any accident previously

evaluated in the Safety Analysis Report?

The change in reactor thermal power will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Equipment that could be affected by power

uprate has been evaluated. No new operating

mode, safety related equipment lineup,

accident scenario, or equipment failure mode

was identified. The full spectrum of accident

considerations defined in the BNP

[Brunswick Nuclear Plant] UFSAR has been

evaluated and no new or different kind of

accident has been identified. Uprate uses

developed technology and applies it within

the capabilities of existing plant equipment

in accordance with existing regulatory

criteria including NRC approved codes,

standards, and methods. General Electric has

designed BWRs [Boiling Water Reactors] of

higher power levels than the uprated power

of any of the currently uprated BWR/4 fleet

and has not identified new power dependent

accidents.

The changes to the Technical

Specifications required to implement power

uprate make little change to the plant’s

configuration. These changes fall into three

major categories. The first includes those

changes resulting from power uprate

parameter changes. These parameter changes,

such as the increase in vessel pressure,

temperature and piping system flows are

minor in nature. The evaluations have shown

the plant is still within its design capabilities

when operating under these conditions. The

changes required as a result of power uprate

will not affect the design function(s) of

currently installed equipment; therefore,

there is no possibility of a new or different

kind of failure mode. The second set of

changes is a result of applying setpoint

methodology to calculate TS Allowable

Values and Normal Trip Setpoints for

instruments that are directly affected by the

parameter changes due to power uprate. By

using CP&L’s methodology, the TS values

were calculated to ensure adequate margin

exists between the analytical limit and the TS

Allowable Value. The third change include

[sic] setpoints that were reconstituted by the

power uprate project. Again, CP&L

methodology was applied and the results

show the setpoints have moved to a more

conservative value. This will reduce the

likelihood of spurious scrams and

unnecessary challenges to safety systems

while ensuring initiation/actuation

equipment continues to function consistent

with existing accident analyses.

3. Does the proposed activity involve a

significant reduction in a margin of safety

defined in the basis of any Operating License

Technical Specification?

Power Uprate will not involve a significant

reduction in a margin of safety. The

bounding events which had been analyzed in

the UFSAR were reevaluated to demonstrate

that power uprate can be implemented

without exceeding any analyzed limit.

Because the applicable safety analysis criteria

and limits are satisfied for power uprate, the

margin of safety associated with the safety

limits and other limits identified in the

Technical Specifications will be maintained.

As discussed in Section 5 of GE Nuclear

Energy’s License Topical Report NEDO31984P ‘‘Generic Evaluations of General

Electric Boiling Water Reactor Power

Uprate,’’ the safety margins prescribed by the

Code of Federal Regulations (CFR) have been

maintained by meeting the appropriate

regulatory criteria. Similarly, the margins

provided by the application of the ASME

design criteria have been maintained. The

Brunswick unique analysis NEDC-32466P

‘‘Power Uprate Safety Analysis Report for

Brunswick Steam Electric Plant Units 1 and

2’’ discusses the effects of power uprate on

safety margins for (1) fuel thermal limits, (2)

design basis accidents and the challenges for

fuel, containment and radiological releases,

(3) transient analysis, (4) non-LOCA

radiological releases, and (5) environmental

consequences. These evaluations conclude

that applicable safety analysis criteria and

limits are satisfied, and thus, the margins of

safety will be maintained.

The changes to the Technical Specification

instrumentation will not involve a reduction

in the margin of safety. The calculations

performed for power uprate have established

an analytical limit and calculated the TS

Allowable Value and Nominal Trip Setpoint

using formal setpoint methodology. This

ensures the instrumentation functional

requirements are met.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403-

3297.

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Eugene V.

Imbro
Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices 25699

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: March

29, 1996

Description of amendment request:

The proposed amendment would revise

the technical specifications (TS) to add

an allowance to complete a TS required

surveillance within 24 hours of

discovery of a missed surveillance in

accordance with the guidance of

Generic Letter (GL) 87-09, ‘‘Sections 3.0

and 4.0 of the Standard Technical

Specifications (STS) on the

Applicability of Limiting Conditions for

Operation and Surveillance

Requirements.’’ The wording specifying

intervals for testing has been changed to

reflect wording consistent the new STS.

Typographical errors in the basis are

also being corrected.

Basis for proposed no significant

hazards consideration determination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed changes do not involve

a significant increase in the probability

or consequences of an accident

previously evaluated.

The proposed changes clarify and

incorporates [sic] NRC guidance for

application of extending or moving

surveillance intervals by plus or minus 25%,

by elimination of restrictive surveillance

interval descriptions that conflict with NRC

guidance, by allowing for an additional 24

hours to perform missed surveillances, and

by providing a defined finite period for the

term ‘‘immediate’’ for Technical

Specification (TS) and Inservice Inspection

(ISI) surveillances. The basis for extending or

moving surveillances, as stated in GL 89-14,

‘‘Line-Item Improvements in Technical

Specifications - Removal of the 3.25 Limit on

Extending Surveillance Intervals,’’ is to

provide plants flexibility for scheduling the

performance of surveillances and to permit

consideration of plant operating conditions

that may not be suitable for conducting a

surveillance at the specified time interval.

Such operating conditions include transient

plant operation or ongoing surveillance or

maintenance activities. Extending

surveillance intervals during plant operation

can result in a benefit to safety when a

scheduled surveillances [sic] is due at a time

that is not suitable for conducting the

scheduled surveillance. NUREG-1431,

‘‘Standard Technical Specifications -

Westinghouse Plants,’’ states ‘‘the 25%

extension does not significantly degrade the

reliability that results from performing the

surveillance at its specified frequency.’’ This

is based on the recognition that the most

probable result of any particular surveillance

being performed is the verification of

conformance with the surveillance

requirements. The basis for the 24 hour delay

period, as stated in the basis for NUREG1431, includes consideration of unit

conditions, adequate planning, availability of

personnel, the time required to perform the

surveillance, the recognition that the most

probable result of any particular surveillance

being performed is the verification of

conformance with the requirements.’’ The

basis for defining the term ‘‘immediate’’ is to

provide guidance to plant personnel for

conducting operability testing of the Steam

Driven Auxiliary Feedwater pump after

extended shutdown periods in order to

minimize plant risks and not pose an unsafe

operational transient during an unstable

plant configuration (i.e., during plant

startup). Since these changes do not affect

plant design, operation, or the manner in

which testing is performed, 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 accident previously

evaluated.

The proposed changes clarify and

incorporates [sic] NRC guidance for

application of extending or moving

surveillance intervals by plus or minus 25%,

by elimination of restrictive surveillance

interval descriptions that conflict with NRC

guidance, by allowing for an additional 24

hours to perform missed surveillances, and

by providing a defined finite period for the

term ‘‘immediate’’ for TS and ISI

surveillances. Since these changes do not

affect plant design, operation, or the manner

in which testing is performed, the proposed

changes do not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

3. The proposed changes do not involve a

significant reduction in the margin of safety.

The changes proposed, with the exception

of allowing an additional 24 hours to

complete missed surveillances, are to clarify

existing surveillance intervals and to provide

more specific and detailed criteria without

changing current surveillance scheduling

methodologies. The NRC has determined that

allowing an additional 24 hours to complete

missed surveillance tests minimizes

additional challenges to plant operations

such that there is a conservative balance

between the risk associated with performing

the surveillance during stable plant

conditions and the risk of imposing a plant

transient due to TS action statements or

changing ‘‘modes’’ of operation. These

extensions are current industry practices

endorsed by the NRC which provide

flexibility for scheduling and performing

surveillances and permit consideration of

plant operating conditions that may not be

suitable for conducting a surveillance at

either the specified time interval or

inadvertently missing the surveillance

interval. The risk to safety is low in contrast

to the alternatives; therefore, the proposed

changes do not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

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: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Eugene V.

Imbro

Commonwealth Edison Company,

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

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of amendment request: April 8,

1996

Description of amendment request:

The proposed amendments would

change various sections of the Technical

Specifications (TS) to reflect the

transition of fuel supplier from Generic

Electric to Siemens Power Corporation

(SPC). The amendments would revise

the definitions and Limiting Conditions

for Operation related to Linear Heat

Generation Rate, Critical Power Ratio,

Maximum Critical Power Ratio, and

Fraction of Limiting Power Density to

incorporate SPC terms and methodology

or to make the TS vendor neutral.

Section 6.0 of the TS would be revised

to include SPC references. The proposed

amendment also adds a requirement to

adjust the Average Planar Linear Heat

Generation Rate when the reactor is in

single loop operation since SPC

methodologies may require this

reduction factor for SPC fuel. The SPC

methodologies to be added to the TS

have previously been approved by the

NRC. The proposed amendment would

also relocate requirements for the

traversing in-core probe system from the

TS to the Core Operating Limits Report

and would upgrade the fuel description

in Section 5.0 as a line item from the

Improved Technical Specifications.

Basis for proposed no significant

hazards consideration determination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The probability of an evaluated accident is

derived from the probabilities of the

individual precursors to that accident. The

consequences of an evaluated accident are

determined by the operability of plant
25700 Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices

systems designed to mitigate those

consequences. Limits will be established

consistent with NRC approved methods to

ensure that fuel performance during normal,

transient, and accident conditions is

acceptable. The proposed Technical

Specifications amendment reflects previously

approved SPC methodology used to analyze

normal operations, including anticipated

operational occurrences (AOOs), and to

determine the potential consequences of

accidents.

Licensing Methods and Models

The proposed amendment is to support

operation with NRC approved fuel and

licensing methods supplied from Siemens

Power Corporation. In accordance with FSAR

Chapter 15, the same accidents and transients

will be analyzed with the new fuel and

methods as were analyzed by GE for GE fuel.

The analysis methods and models are NRC

approved (Note the mixed core treatment of

critical power ratio is being addressed under

separate correspondence). These approved

methods and models are used to determine

the fuel thermal limits. Traversing In-core

Probe (TIP) uncertainty are assumptions in

the approved Siemens core monitoring

methodologies. The SPC core monitoring

code enables the site to monitor keff as well

as rod density to perform the reactivity

anomaly surveillance. This is consistent with

GE methodology. Therefore, the change in

licensing analysis methods and models does

not significantly increase the probability of

an accident or the consequences of an

accident previously identified. The support

systems for minimizing the consequences of

transients and accidents are not affected by

the proposed amendment.

New Fuel Design

The use of ATRIUM 9B fuel at LaSalle does

not involve a significant increase in the

probability or consequences of any accident

previously evaluated in the FSAR. The

ATRIUM-9B fuel is generically approved for

use as a reload BWR fuel type. (See Boiling

Water Reactor Licensing Methodology

Summary, Siemens Power Corporation, EMF94-217(NP)). Limiting postulated occurrences

and normal operation have been analyzed

using NRC-approved methods for the

ATRIUM 9B fuel design to ensure that safety

limits are protected and that acceptable

transient and accident performance is

maintained.

The reload fuel has no adverse impact on

the performance of in-core neutron flux

instrumentation or control rod drive

response. The ATRIUM-9B fuel design will

not adversely affect performance of neutron

instrumentation nor will it adversely affect

the movement of control blades. The exterior

dimensions of the ATRIUM-9B fuel assembly

are essentially identical to the GE9B; the

ATRIUM-9B fuel assembly for LaSalle uses a

standard fuel channel and normal control

cell positioning (i.e., no offset). Thus, no

adverse interactions with the adjacent control

blade and nuclear instrumentation are

anticipated. Additionally, given the above

mentioned overall envelope similarities, no

problems are anticipated with other station

equipment such as the fuel storage racks, the

new fuel inspection stand and the spent fuel

pool fuel preparation machine.

The ATRIUM 9B design is neutronically

compatible with the existing fuel types and

core components in the LaSalle core. SPC

tests have demonstrated that the ATRIUM-9B

fuel design is hydraulically compatible with

the GE9 fuel. The bundle pressure drop

characteristics of the ATRIUM 9B bundle are

similar to those of the GE9 fuel design, hence

core thermal-hydraulic stability

characteristics are not adversely affected by

the ATRIUM 9B design.

An evaluation of the Emergency

Procedures is being performed to ensure that

the use of the ATRIUM-9B fuel at LaSalle

does not alter any assumptions previously

made in evaluating the radiological

consequences of an accident at LaSalle

Station.

Methods approved by the NRC are being

used in the evaluation of fuel performance

during normal and abnormal operating

conditions. The ComEd and SPC methods to

be used for the cycle specific transient

analyses have been previously NRC

approved. The exception is the mixed core

treatment of critical power ratio, which is

being addressed under separate

correspondence.

The description of the fuel is expanded to

be consistent with NUREG-1434. The

description of the fuel materials, lead test

assembly use, and stating that designs must

have been analyzed with NRC Staff approved

codes does not change existing methods; it

only describes them.

Review of the above concludes that the

probability of occurrence and the

consequences of an accident previously

evaluated in the safety analysis report have

not been significantly increased.

* * * * *

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated:

Creation of the possibility of a new or

different kind of accident would require the

creation of one or more new precursors of

that accident. New accident precursors may

be created by modifications of the plant

configuration, including changes in

allowable modes of operation.

Licensing Methods and Models

The proposed Technical Specification

amendment reflects previously approved SPC

methodology used to analyze normal

operations, including AOOs, and to

determine the potential consequences of

accidents. As stated above, the proposed

changes do not permit modes of reactor

operation which differ from those currently

permitted.

New Fuel Design

The basic design concept of a 9x9 fuel pin

array with an internal water box has been

used in various lead assembly programs and

in reload quantities in Europe since 1986.

WNP-2 has loaded reload quantities since

1991. Approximately 650 water box

assemblies have been irradiated in the United

States through 1995, with a substantially

higher number being irradiated overseas. The

NRC has reviewed and approved the

ATRIUM-9B fuel design. (See Boiling Water

Reactor Licensing Methodology Summary,

Siemens Power Corporation, EMF-94-

217(NP)). The similarities in fuel design and

operation indicate there would be no

expectation of introducing new or different

types of accidents than have been considered

for the existing fuel. Therefore, the use of

ATRIUM-9B fuel at LaSalle does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

* * * * *

3. Involve a significant reduction in the

margin of safety for the following reasons:

The existing margin to safety is provided

by the existing acceptance criteria (e.g.,

10CFR50.46 limits). The proposed Technical

Specification amendment reflects previously

approved SPC methodology used to

demonstrate that the existing acceptance

criteria are satisfied. The revised

methodology has been previously reviewed

and approved by the USNRC for application

to reload cores of GE BWRs. References for

the Licensing Topical Reports which

document this methodology, and include the

Safety Evaluation Reports prepared by the

USNRC, are added to the Reference section

of the Technical Specifications as part of this

amendment.

Licensing Methods and Models

The proposed amendment does not involve

changes to the existing operability criteria.

NRC approved methods and established

limits (implemented in the Core Operating

Limits Report) ensure acceptable margin is

maintained. The ComEd and SPC reload

methodologies for the ATRIUM-9B reload

design are consistent with the Technical

Specification Bases. The Limiting Conditions

for Operation are taken into consideration

while performing the cycle specific and

generic reload safety analyses. NRC approved

methods are listed in Specification 6.0 of the

Technical Specifications.

Analyses performed with NRC-approved

methodology have demonstrated that fuel

design and licensing criteria will be met

during normal and abnormal operating

conditions. Therefore, there is not a

significant reduction in the margin of safety.

New Fuel Design

The exterior dimensions of the ATRIUM9B fuel assembly are essentially identical to

the GE9B; the ATRIUM-9B fuel assembly for

LaSalle uses a standard fuel channel and

normal control cell positioning; i.e., no offset.

Thus, no adverse interactions with the

adjacent control blade and nuclear

instrumentation are anticipated. The change

does not adversely impact equipment

important to safety and, therefore does not

reduce the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One
Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices 25701

First National Plaza, Chicago, Illinois

60603

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: April 9,

1996

Description of amendment request:

The proposed amendments would

eliminate the automatic reactor scram

function and the group 1 and 3 isolation

valve closure functions associated with

the Main Steam Line Radiation

Monitoring (MSLRM) system high

radiation setpoint. Elimination of these

functions will eliminate potential

spurious scrams and isolations caused

by increased main steam line radiation

levels during hydrogen injection. The

licensee also proposes to raise the

MSLRM system alarm setpoints which

are not part of the Technical

Specifications to include increased

background radiation during hydrogen

injection. The proposed amendment

would also delete the surveillance

requirements for the associated

instruments.

Basis for proposed no significant

hazards consideration determination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1) Involve a significant increase in the

probability or consequences of an accident

previously evaluated because:

Redefining the full power radiation

background, thus changing the MSLRM

alarm setpoint, does not change the

probability of occurrence of any accident

which has been postulated and analyzed in

the UFSAR, but will reduce the probability

of the inadvertent MSIV closure transient

which is an analyzed transient in the UFSAR.

It does not change the probability of

malfunction of any equipment important to

safety associated with [loss of coolant

accident] LOCA, fuel handling accident or

[control rod drop accident] CRDA. It also

does not change the resultant offsite

radiological dose from the bounding design

basis CRDA. This is based upon all

radioactivity, resulting from the design basis

CRDA, going to the condenser

instantaneously (or independent of the actual

MSLRM setpoint) in the offsite dose

calculation.

The elimination of reactor scram and

isolation of MSIVs, isolation of main steam

line drain valves and reactor water sample

line valves, associated with the MSLRM

system actuation do not introduce, mitigate,

or reduce the probability of any design basis

accident, or any accident, evaluated in the

UFSAR. The topical report NEDO-31400A

has shown that there is essentially no

reasonable radiological consequence benefit

in a design basis CRDA of retaining the

MSLRM associated reactor scram and MSIV

isolation function. In addition, the

probability of inadvertent scram and

isolation is reduced. The proposed change

will not adversely impact the operation of the

[reactor protection system] RPS or [primary

containment isolation system] PCIS with

respect to performing its other intended

safety functions. The proposed change will

not affect the operation of other plant

systems or equipment important to safety.

The consequences of eliminating the

automatic closure of the main steam line

drain isolation valves and reactor

recirculation water sample line isolation

valves along with the MSIVs has been

evaluated to be negligible additions to the

CRDA doses. A [LaSalle County Station]

LSCS unique analysis has demonstrated that

the radiological doses as a result of design

basis CRDA are acceptable.

The MSLRM system high radiation trip

was intended to function in response to a

CRDA which has been previously evaluated.

No credit for MSIV closure was taken in the

CRDA analysis since it postulates that all the

radioactive material assumed to be released

from the fuel is transported to the main

condenser prior to MSIV closure.

Furthermore, the probability of a fuel failure

is independent of the operation of the

MSLRM system.

By eliminating the MSLRM induced MSIV

closure, the Offgas system can be utilized to

reduce potential offsite doses after a CRDA.

The [mechanical vacuum pump] MVP is

tripped no later than 15 minutes of a Hi-Hi

radiation alarm but analytically results in

acceptable offsite doses.

Thus the proposed amendment will not

increase the probability of any accident

previously evaluated, and the elimination of

the MSLRM isolation signal for MSIVs and

other small containment valves will not

significantly increase the consequences of a

CRDA as previously evaluated.

2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated because:

Redefining the full power radiation

background, thus changing the actual

MSLRM alarm setpoint, does not alter the

configuration of the plant. It does not revise

any logic or function of the MSLRM trip

channels or add, replace, or delete any

equipment important to safety. Therefore it

does not introduce any new failure modes or

create any possibility of a new accident

which may challenge safety to the public and

has not been previously analyzed. It also

does not involve any equipment which either

has not been evaluated previously, or may

have any safety consequences to the public.

The proposed Technical Specification

changes involve eliminating the MSLRM

system high radiation trip function for

initiating an automatic reactor scram, and

automatic isolations. The proposed changes

will not affect the operation of other plant

systems or equipment important to safety.

The MSLRM system will continue to initiate

alarms as before. Plant procedures will be in

place to take appropriate mitigative measures

in response to a high alarm.

The isolation and reactor scram functions

associated with the MSLRM system actuation

were originally intended to mitigate, not

prevent, a potential accident scenario such as

a CRDA or gross fuel failure event. Adding

or removing an electronic signal, such as the

one from the MSLRM system, does not

change system or hardware design within the

reactor vessel pressure boundary, and

therefore will not create the possibility of a

new or different kind of accident from those

evaluated in the UFSAR like a LOCA or

CRDA during power operation. It also does

not create the possibility of a new or different

kind of accident outside the reactor vessel

pressure boundary from those evaluated in

the UFSAR, such as a LOCA or Fuel

Handling Accident. Removing the isolation

signal also reduces the probability of

inadvertent scram and isolation.

Therefore the proposed amendment will

not create the possibility of a new or different

kind of accident from any accident

previously analyzed.

3) Involve a significant reduction in the

margin of safety because:

The current MSLRM trip Hi-Hi alarm

setpoint (about 4 R/hour with full power

background at 1.3 R/hour) is at 3 times the

full power radiation background. As

indicated in the plant unique analytical

result for LSCS, the radiological reading at

the MSLRMs for design basis CRDA is

equivalent to over 1200 times the normal full

power radiation background (1600 R/hour

divided by 1.3 R/hour), or 150 times the full

power radiation background during peak

HWC environment (since the radiation

background is 8 times the normal

background). Thus the safety margin was

very large, and would still be quite large with

the HWC background factored into the

MSLRM actuation setpoint (3 x 8 x 1.3 =

about 50). The Hi alarm setpoint of 1.5 times

full power background likewise will have a

higher safety margin. Thus there is basically

no adverse consequence to the margin of

safety in the basis for the LaSalle technical

specifications.

The proposed Technical Specification

changes to eliminate the MSLRM system high

radiation trip function for initiating an

automatic reactor scram, and automatic

closure of the MSIVs, main steam line drain

isolation valves, and reactor recirculation

water sample line isolation valves do not

cause radiological dose consequences to

exceed the limit established by SRP 15.4.9.

Per NEDO-31400A, the elimination of

MSLRM trip/scram signal will result in the

reduction of potential inadvertent scrams,

unnecessary safety-related actuations, undue

vessel isolation, and duty challenges during

normal plant operation. These can be

interpreted to be a potential reduction in core

damage frequency, which translates to an

improvement in the margin of safety.

Thus the margin of safety as defined in the

basis of the technical specifications is

essentially unaffected, and is therefore

acceptable.

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
25702 Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: 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: April 16,

1996

Description of amendment request:

The proposed amendments would

eliminate the Technical Specification

requirement to perform response time

testing for selected instruments. The

instruments affected are the sensors for

selected reactor protection system

instrumentation, main steam isolation

actuation instrumentation, and all

sensors for emergency core cooling

system (ECCS) actuation

instrumentation. The proposed changes

are supported by analyses performed by

the Boiling Water Reactor Owners’

Group as documented in NEDO-32291-

A which was approved by the NRC for

use in license amendment applications.

Basis for proposed no significant

hazards consideration determination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1) Involve a significant increase in the

probability or consequences of an

accident previously evaluated because:

The purpose of the proposed Technical

Specification (TS) change is to eliminate

response time testing requirements for

selected components in the Reactor

Protection System (RPS), Isolation Actuation

instrumentation and Emergency Core Cooling

System (ECCS) actuation instrumentation.

The Boiling Water Reactor Owners’ Group

(BWROG) has completed an evaluation

which demonstrates that response time

testing is redundant to the other TS-required

testing. These other tests, in conjunction with

actions taken in response to NRC Bulletin 90-

01, ‘‘Loss of Fill-Oil in Transmitters

Manufactured by Rosemount,’’ and

Supplement 1, are sufficient to identify

failure modes or degradations in instrument

response time and ensure operation of the

associated systems within acceptable limits.

There are no known failure modes that can

be detected by response time testing that

cannot also be detected by the other TSrequired testing. This evaluation was

documented in NEDO-32291-A, ‘‘System

Analyses for the Elimination of Selected

Response Time Testing Requirements,’’ dated

October 1995. LaSalle County Station,

LaSalle, has confirmed the applicability of

this evaluation to LaSalle. In addition,

LaSalle will complete the actions identified

in the NRC staffs safety evaluation of NEDO32291-A.

Because of the continued application of

other existing TS-required tests such as

channel calibrations, channel checks,

channel functional tests, and logic system

functional tests, the response time of these

systems will be maintained within the

acceptance limits assumed in plant safety

analyses and required for successful

mitigation of an initiating event. The

proposed changes do not affect the capability

of the associated systems to perform their

intended function within their required

response time, nor do the proposed changes

themselves affect the operation of any

equipment. As a result, LaSalle has

concluded that the proposed changes do not

involve a significant increase in the

probability or the consequences of an

accident previously evaluated.

2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated because:

The proposed changes only apply to the

testing requirements for the components

identified above and do not result in any

physical change to these or other components

or their operation. As a result no new failure

modes are introduced. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3) Involve a significant reduction in the

margin of safety because:

The current TS-required response times are

based on the maximum allowable values

assumed in the plant safety analyses. These

analyses conservatively establish the margin

of safety. As described above, the proposed

changes do not affect the capability of the

associated systems to perform their intended

function within the allowed response time

used as the basis for the plant safety analyses.

The potential failure modes for the

components within the scope of this request

were evaluated for impact on instrument

response time. This evaluation confirmed

that, with the exception of loss of fill-oil of

Rosemount transmitters, the remaining TSrequired testing is sufficient to identify

failure modes or degradations in instrument

response times and ensure that operation of

the applicable instrumentation is within

acceptable limits. The actions taken in

response to NRC Bulletin 90-01 and

Supplement 1 are adequate to identify loss of

fill-oil failures of Rosemount transmitters. As

a result, it has been concluded that plant and

system response to an initiating event will

remain in compliance with the assumptions

of the safety analysis.

Further, although not explicitly evaluated,

the proposed changes will provide an

improvement to plant safety and operation by

the following:

a. Reducing the time safety systems are

unavailable,

b. Reducing the potential for safety system

actuations,

c. Reducing plant shutdown risk,

d. Limiting radiation exposure to plant

personnel, and

e. Eliminating the diversion of key

personnel resources to conduct unnecessary

testing.

Therefore, LaSalle has concluded that this

request will not significantly reduce the

margin of safety, and may actually cause an

increase in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: Robert A. Capra

Duke Power Company, et al., Docket

Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of amendment request:

November 2, 1994

Description of amendment request:

The proposed amendments would

delete the content of Appendix B,

‘‘Environmental Protection Plan’’

(nonradiological), and modify License

Condition 2.C.(2) to delete that portion

which refers to the Environmental

Protection 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 presented

below:

1. [The proposed amendments would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated]:

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.

2. [The proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated]:

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.

3. [The proposed amendments would not

involve a significant reduction in a margin of

safety]:
Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices 25703

The proposed revision will not reduce a

margin of safety because it is administrative

in nature and will not [a]ffect the margin of

safety as defined in the basis for any

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: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Attorney for licensee: Mr. Albert Carr,

Duke Power Company, 422 South

Church Street, Charlotte, North Carolina

28242

NRC Project Director: Herbert N.

Berkow

Duquesne Light Company, et al., Docket

No. 50-412, Beaver Valley Power

Station, Unit 2, Shippingport,

Pennsylvania

Date of amendment request: April 29,

1996

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 5.3.1 to

allow the use of ZIRCO as an alternate

zirconium-based fuel rod material and

remove the word clad since it has been

eliminated from the text of the NRC’s

improved Standard Technical

Specifications (NUREG-1431). Limited

substitution of fuel rods by ZIRCO filler

rods would also be permitted. The

proposed amendment would revise Note

2 on TS Table 3.9-1 to specify that the

maximum burnup in the peak fuel rod

in a fuel assembly stored in Region 2

spent fuel racks should not exceed the

NRC-approved limit for WCAP-12610

rather than the current maximum

burnup limit of 60 GWD/MTU.

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 methodologies used in the accident

analyses remain unchanged. The proposed

changes do not change or alter the design

assumptions for the systems or components

used to mitigate the consequences of an

accident. Use of ZIRLO fuel rod material does

not adversely affect fuel performance or

impact nuclear design methodology.

Therefore, accident analysis results are not

impacted.

The operating limits will not be changed

and the analysis methods to demonstrate

operation within the limits will remain in

accordance with NRC approved

methodologies. Other than the changes to the

fuel assemblies, there are no physical

changes to the plant associated with this

technical specification change. A safety

analysis will continue to be performed for

each cycle to demonstrate compliance with

all fuel safety design bases.

VANTAGE 5 fuel assemblies with ZIRLO

fuel rods meet the same fuel assembly and

fuel rod design bases as other VANTAGE 5

fuel assemblies. In addition, the 10 CFR

50.46 criteria are applied to the ZIRLO fuel

rods. The use of these fuel assemblies will

not result in a change to the reload design

and safety analysis limits. Since the original

design criteria are met, the ZIRLO fuel rods

will not be an initiator for any new accident.

The fuel rod material is similar in chemical

composition and has similar physical and

mechanical properties as Zircaloy-4. Thus,

the fuel rod integrity is maintained and the

structural integrity of the fuel assembly is not

affected. ZIRLO improves corrosion

performance and dimensional stability. No

concerns have been identified with respect to

the use of an assembly containing a

combination of Zircaloy-4 and ZIRLO fuel

rods.

The dose predictions in the safety analyses

are not sensitive to the fuel rod material

used; therefore, the radiological

consequences of accidents previously

evaluated in the safety analysis remain valid.

A reload analysis is completed for each cycle,

in accordance with NRC approved

methodologies. 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.

VANTAGE 5 fuel assemblies with ZIRLO

fuel rods satisfy the same design bases as

those used for other VANTAGE 5 fuel

assemblies. All design and performance

criteria continue to be met and no new

failure mechanisms have been identified. The

ZIRLO fuel rod material offers improved

corrosion resistance and structural integrity.

The proposed changes do not affect the

design or operation of any system or

component in the plant. The safety functions

of the related structures, systems, or

components are not changed in any manner,

nor is the reliability of any structure, system,

or component reduced. The changes do not

affect the manner by which the facility is

operated and do not change any facility

design feature, structure, or system. No new

or different type of equipment will be

installed. Since there is no change to the

facility or operating procedures, and the

safety functions and reliability of structures,

systems, or components are not affected, 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 use of Zircaloy-4, ZIRLO, or stainless

steal filler rods in fuel assemblies will not

involve a significant reduction in the margin

of safety because analyses using NRC

approved methodology will be performed for

each configuration to demonstrate continued

operation within the limits that assure

acceptable plant response to accidents and

transients. These analyses will be performed

using NRC approved methods that have been

approved for application to the fuel

configuration.

Use of ZIRLO as fuel rod material does not

change the VANTAGE 5 reload design and

safety analysis limits. The use of these fuel

assemblies will take into consideration the

normal core operating conditions allowed in

the technical specifications. For each reload

core, the fuel assemblies will be evaluated

using NRC approved reload design methods,

including consideration of the core physics

analysis peaking factors and core average

linear heat rate effects.

Based on the above, it is concluded that the

proposed license amendment request does

not result in a significant reduction in margin

with respect to plant safety as defined in the

UFSAR [Updated Final Safety Analysis

Report] or any plant technical specification

BASES.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 1500l.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz

Entergy Operations, Inc., Docket Nos.

50-313 and 50-368, Arkansas Nuclear

One, Unit Nos. 1 and 2 (ANO-1&2),

Pope County, Arkansas

Date of amendment request: May 2,

1996

Description of amendment request:

The proposed technical specification

amendments would extend the allowed

outage times for emergency diesel

generators at Arkansas Nuclear One,

Units 1 and 2 to 7 days with an

additional, once per refueling cycle

extension of 7 more days for each

machine.

Basis for proposed no significant

hazards consideration determination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Criterion 1 - Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

The emergency diesel generators (EDGs)

are backup alternating current power sources
25704 Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices

designed to power essential safety systems in

the event of a loss of offsite power. The EDGs

are not accident initiators in any accident

previously evaluated. Probabilistic Safety

Analysis (PSA) methods were utilized in

order to fully evaluate the EDG allowed

outage time (AOT) extension proposed in this

submittal. The results of these analyses

indicate there is not a significant increase in

the probability of an accident previously

evaluated. Therefore, this change does not

involve an increase in the probability of an

accident previously evaluated.

The EDGs provide backup power to

components that mitigate the consequences

of accidents. The current TSs allow for an

EDG to be removed from service for an AOT.

The proposed amendment extends the

current AOT for an EDG. The proposed

change does not allow any more equipment

to be removed from service at one time. The

proposed changes to the AOTs do not affect

any of the assumptions used in deterministic

safety analysis. By extending the EDG AOT,

the consequences of an accident previously

evaluated will remain unchanged.

The proposed change removes redundant

requirements associated with an inoperable

emergency power supply from the TS for the

pressurizer proportional heaters. The

operability requirements for emergency

power supplies and actions to be taken if an

EDG is inoperable are already addressed in

the ANO-2 TS 3.8.1.1.

The associated changes that remove the

requirements to test the EDGs if one or both

offsite power supplies are inoperable, for an

inoperable station battery, for an inoperable

component in the two ESF electrical

distribution systems, the accelerated testing

requirements of the EDGs, and the daily

testing requirements for the operable EDGs

improve the reliability for the operable EDGs

by reducing the number of unnecessary starts

and stops. By improving the EDG reliability,

this change will not increase the

consequences of the accidents previously

evaluated.

The other changes in this submittal

associated with the bases are considered

administrative in nature and have no effect

on the consequences of an accident

previously evaluated.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

This proposed change does not alter the

design, configuration, or method of operation

of the plant. Therefore, this change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in the Margin of Safety.

The proposed changes do not affect the

Technical Specification limiting conditions

for operation or their bases which support

the deterministic analyses used to establish

the margin of safety.

Calculations performed to analyze the

change in risk based on these changes

produced acceptable values which are

included in the tables located in the

description of changes section. These

calculated changes in risk fall well within

that which is normally considered

acceptable. When the additional benefit of

maintaining the Emergency Diesel Generators

available during shutdown cooling

operations associated with refueling outages

in considered, the overall change in risk is

further reduced.

The remaining proposed changes are either

associated with increasing EDG reliability or

considered administrative in nature.

Therefore, this change does not involve a

significant reduction in the margin of safety

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the 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: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., Washington, DC

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., et al., Docket

No. 50-416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi

Date of amendment request:

November 20, 1995, as supplemented by

the letter dated December 15, 1995.

Description of amendment request:

The licensee has proposed to revise the

Grand Gulf Nuclear Station (GGNS),

Unit 1, Technical Specifications (TSs) as

follows for the drywell, the drywell

airlock, and the drywell isolation

valves:

1. For the drywell in Limiting

Condition of Operation (LCO) 3.6.5.1,

the surveillance frequency interval for

the drywell bypass test in Surveillance

Requirement (SR) 3.6.5.1.1 would be

increased from 18 months to 10 years.

For this interval change, an increased

testing frequency would be required if

bypass performance degrades (i.e., the

leakage is greater than the limit for two

consecutive tests) and the application of

SR 3.0.2, the allowance to extend the

surveillance interval by 25 percent,

would be restricted to 12 months on the

10-year interval. This includes deleting

the Note in SR 3.6.5.1.1.

2. For the drywell airlock in LCO

3.6.5.2, the following changes are

requested: (a) the leak rate SR 3.6.5.2.2

would be transferred from the airlock

LCO (3.6.5.2) to SR 3.6.5.1.3 in the

drywell LCO (3.6.5.1), (b) the

requirement in SR 3.6.5.2.2 for the air

lock to meet a specific overall leakage

limit would be deleted, (c) the Note in

SR 3.6.5.2.2 that stated that an

inoperable air lock door does not

invalidate the previous air lock leakage

test would be deleted, (d) the test

pressure for the air lock leakage test in

SR 3.6.5.2.2 would be reduced from 11.5

psig to 3 psid, and (e) the surveillance

frequency interval for the air lock

leakage and interlock testing, required

in SRs 3.6.5.2.1 and 3.6.5.2.2, would be

increased from 18 months to 24 months.

3. For the drywell airlock in LCO

3.6.5.2 and the drywell isolation valves

in LCO 3.6.5.3, the Action Notes, which

identify that the actions required by

drywell LCO 3.6.5.1 must be taken

when the drywell bypass leakage limit

is not met, would be deleted. Action C.1

of LCO 3.6.5.2 and its associated

completion time would also be

deleted.There would also be changes to

the Bases of the TSs for the above LCOs

and SRs, based on the proposed

changes.

Basis for proposed no significant

hazards consideration determination:

The amendment request dated

November 20, 1995, applied to both the

Grand Gulf Nuclear Station (GGNS) and

the River Bend Station (RSB); however,

not all of the proposed amendments

apply to GGNS. This Notice only

discusses the amendment request for

GGNS. The reference below to proposed

amendments which do not apply to

GGNS are marked by ‘‘[....]’’.

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration in its application dated

November 20, 1995, which is presented

below:

Entergy Operations, Inc. proposes to

change the current Grand Gulf Nuclear

Station (GGNS) [....] Technical Specifications.

The specific proposed changes are:

1. The Surveillance Frequency [interval]

for the drywell bypass test is changed

[increased] from 18 months to 10 years with

an increased testing frequency required if

performance degrades.

2. The following changes are requested for

the drywell air lock testing: (a) the leakage

rate surveillance is moved from the air lock

Limiting Condition for Operation (LCO) to

the drywell LCO, (b) the requirement for the

air lock to meet a specific overall leakage

limit is deleted, (c) the Note that an

inoperable air lock door does not invalidate

the previous air lock leakage test is deleted,

(d) the GGNS test pressure for the air lock

leakage test is changed [reduced] from 11.5

psig to 3 psid, [...,] and ([e]) the Surveillance

Frequency [interval] for the air lock leakage

test and interlock test is changed [increased]

from 18 months to 24 months.

3. The Actions Notes in the drywell air

lock LCO and the drywell isolation valve

LCO that identifies that the Actions required
Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices 25705

by the drywell LCO must be taken when the

drywell bypass leakage limit is not met is

deleted. [Action C.1 of LCO 3.6.5.2 and its

associated completion time would also be

deleted.]

[4. ...]

The Commission has provided standards

for determining whether a no significant

hazards consideration exists as stated in 10

CFR 50.92(c). The proposed changes involve

the withdrawal of operating restrictions

previously imposed because acceptable

operation of the Mark III primary

containment design had not been

demonstrated at the time of licensing. As

published in the Federal Register regarding

no significant hazards consideration criteria,

granting of a relief, based upon

demonstration of acceptable operation from

an operating restriction that was imposed

because acceptable operation had not yet

been demonstrated does not involve a

significant hazards consideration (Ref. 48 FR

14870). Furthermore, a proposed amendment

to an operating license involves no

significant hazards consideration if 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.

Entergy Operations, Inc. has evaluated the

no significant hazards consideration in its

request for this license amendment, even

though the above-mentioned criterion is

satisfied by this proposal. In accordance with

10 CFR 50.91(a), Entergy Operations, Inc. is

providing the analysis of the proposed

amendment against the three standards in 10

CFR 50.92(c). A description of the no

significant hazards consideration

determination follows:

I. The proposed change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

The requested changes are either

administrative changes which clarify the

format of the requirement or change the

requirement to match the design bases of the

plant, a change which relocates the

requirement to the Technical Specification

Bases, or a change in [the] surveillance

interval. Each of these types of change are

discussed below:

1. The administrative changes clarify the

format of the requirement or change the

requirement to match the design bases of the

plant. Clarifying [the] administrative format

of the Technical Specifications does not

result in any changes to the Technical

Specification requirements and, as a result,

does not involve a significant increase in the

probability or consequences of an accident

previously evaluated. Also, changing the

requirements of the Technical Specifications

to more closely match the design bases of the

plant will continue to assure that the plant

will respond as assumed in the accident

analyses and, as a result, does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed changes relocate

information to the Technical Specification

Bases. In the Technical Specifications Bases

the relocated information will be maintained

in accordance with 10 CFR 50.59 and subject

to the change control provisions in Chapter

5 of Technical Specifications. Since any

changes to the Technical Specifications Bases

will be evaluated per the requirements of 10

CFR 50.59, no increase (significant or

insignificant) in the probability or

consequences of an accident previously

evaluated will be allowed. Therefore, this

change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

3. The proposed changes in frequency for

the drywell bypass leakage and drywell air

lock surveillances will continue to ensure

that no paths exist through passive drywell

boundary components that would permit

gross leakage from the drywell to the primary

containment air space and result in

bypassing the primary containment pressuresuppression feature beyond the design basis

limit. The Mark III primary containment

system satisfies General Design Criterion 16

of Appendix A to 10 CFR Part 50. Maximum

drywell bypass leakage was determined

previously by reviewing the full range of

postulated primary system break sizes. The

limiting case was a primary system small

break loss of coolant accident (LOCA) and

yielded a design allowable drywell bypass

leakage rate limit of approximately 35,000

scfm for GGNS [....]. The Technical

Specifications acceptable limit for the bypass

leakage following a surveillance is less than

10% of this design basis value. The most

recent bypass leakage value was

approximately 2.5% for GGNS [....] of the

design allowable leakage rate limit for the

limiting event. EOI [Entergy Operations, Inc.]

is committed to maintaining programmatic

and oversight controls that ensure that

drywell bypass leakage remains a small

fraction of the design allowable leakage limit.

The drywell is typically exposed to

essentially 0 psig during normal plant

operation and 3 psig during drywell bypass

leak rate testing. These pressures are

considerably lower than the structural

integrity test pressure and are less likely to

initiate a crack or cause an existing crack to

grow. Visual inspections of the accessible

drywell surfaces that have been performed

since the structural integrity tests have not

revealed the presence of additional cracking

or other abnormalities. Therefore, additional

cracking of the drywell structure is not

expected due to testing or operation and,

similar to the justification for the ten year 10

CFR 50 Appendix J Type A test interval, it

is not considered credible for the passive

drywell structure to begin to leak sufficiently

to impact the design drywell bypass leakage

limit.

The primary containment’s ability to

perform its safety function is fairly

insensitive to the amount of drywell leakage,

thereby providing a margin to loss of the

drywell safety function that is not normally

available for systems. This insensitivity is

demonstrated by the extremely high limiting

event design basis allowable leakage for the

drywell (e.g., 35,000 scfm for GGNS [....]).

The limiting leakage is almost an order of

magnitude higher for other events.

Additionally, an even higher allowable

leakage can be realistically accommodated by

the primary containment due to the margins

in the containment design. Because of the

margins available, it will take valves in

multiple penetration flow paths leaking

excessively to cause the primary containment

to fail as a result of overpressurization, the

probability that drywell isolation valve

leakage will result in primary containment

failure due to excessive drywell leakage is

not considered significant and this drywell/

primary containment failure mode is not

considered credible.

The proposed Technical Specification

changes have no significant impact on the

GGNS Individual Plant Examination (IPE)

[....] conducted per NRC Generic Letter 88-20.

The IPEs considered overpressurization

failure of primary containment as part of the

primary containment performance

assessment. Due to the magnitude of

acceptable drywell leakage and the extremely

low probabilities of achieving such leakage,

primary containment failure due to

preexisting excessive drywell leakage was

considered a non significant contributor to

primary containment failure. Primary

containment overpressurization failure can

occur with or without preexisting excessive

drywell leakage in a severe accident. This is

due to physical phenomena associated with

potentially extreme environmental

conditions inside primary containment

following a severe accident. However, the

calculated frequency of such extreme

conditions is very small. The proposed

changes do not impact the IPE evaluated

phenomena causing primary containment

overpressurization failure nor significantly

increase the probability that the drywell has

preexisting excessive leakage and therefore

would not contribute to these accident

scenarios.

For the reasons discussed above, the

proposed changes do not have any significant

risk impact to accidents previously evaluated

and do not significantly increase the

consequences of an accident previously

evaluated. Additionally, drywell leakage is

not the initiator of any accident evaluated;

therefore, changes in the frequency of the

surveillance for drywell leakage does not

increase the probability of any accident

evaluated.

Therefore, the proposed changes do not

significantly increase the probability or

consequences of an accident previously

evaluated.

II. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The requested changes are either

administrative changes which clarify the

format of the requirement or change the

requirement to match the design bases of the

plant, a change which relocates the

requirement to the Technical Specification

Bases, or a change in surveillance interval.

Each of these types of change are discussed

below:

1. The administrative changes in the

Technical Specification requirements do not
25706 Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices

involve a physical alteration of the plant (no

new or different type of equipment will be

installed) nor does it change the methods

governing normal plant operation. Thus, this

change does not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

2. The proposed relocation of requirements

does not involve a physical alteration of the

plant (no new or different type of equipment

will be installed) nor does it change the

methods governing normal plant operation.

The proposed change will not impose or

eliminate any requirements. Adequate

control of the information will be maintained

in the Technical Specification Bases. Thus,

the change proposed does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed change modifies the

surveillance frequency for drywell bypass

leakage and drywell air lock surveillances.

The changes only impact the test frequency

and do not result in any change in the

response of the equipment to an accident.

The changes do not alter equipment design

or capabilities. The changes do not present

any new or additional failure mechanisms.

The drywell is passive in nature and the

surveillance will continue to verify that its

integrity has not deteriorated. Therefore, the

proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

III. The proposed change does not involve

a significant reduction in a margin of safety.

The requested changes are either

administrative changes which clarify the

format of the requirement or change the

requirement to match the design bases of the

plant, a change which relocates the

requirement to the Technical Specification

Bases, or a change in surveillance interval.

Each of these types of changes are discussed

below:

1. The administrative changes in the

Technical Specification requirements do not

involve a physical alteration of the plant (no

new or different type of equipment will be

installed) nor does it change the methods

governing normal plant operation. Thus, this

change does not cause a significant reduction

in the margin of safety.

2. The relocation of requirements will not

reduce a margin of safety because it has no

impact on any safety analysis assumptions.

In addition, the requirements to be

transferred from the Technical Specifications

to the Technical Specifications Bases are the

same as the existing Technical

Specifications. Since any future changes to

these requirements in the Technical

Specifications Bases will be evaluated per the

requirements of 10 CFR 50.59, no reduction

(significant or insignificant) in a margin of

safety will be allowed.

3. The proposed change modifies the

surveillance frequency for drywell bypass

leakage and associated air lock surveillances.

Reliability of drywell integrity is evidenced

by the measured leakage rate during past

drywell bypass leakage surveillances.

Appropriate design basis assumptions will be

upheld, even when combined with the

complementary bypass leakage surveillances

as proposed. Drywell integrity will continue

to be tested by means of the proposed

periodic drywell bypass leakage test,

performance of the drywell air lock door

latching and interlock mechanism

surveillance, and performance of additional

surveillances including exercising of drywell

isolation valves. The combination of these

surveillances will provide adequate

assurance that drywell bypass leakage will

not exceed the design basis limit. Margins of

safety would not be reduced unless leakage

rates exceeded the design allowable drywell

bypass leakage limit. Therefore, the proposed

change does not cause a significant reduction

in the margin of safety.

Therefore, the proposed changes do not

cause 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: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., 12th Floor,

Washington, DC 20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations Inc., Docket No. 50-

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: August

11, 1995, as supplemented by letter

dated February 12, 1996.

Description of amendment request:

The proposed change will reduce the

minimum reactor coolant cold leg

temperature from 544 Degrees F to 541

degrees F in Technical Specification

Section 3.2.6, ‘‘Reactor Coolant Cold Leg

Temperature.’’

Basis for proposed no significant

hazards consideration determination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed change involves a 3°F

reduction in the minimum core inlet

temperature. This change will not have any

impact on the probability of occurrence of

any accident documented in the FSAR.

The impact of this change on the

consequences of events documented in the

FSAR has been evaluated. The evaluation

demonstrated that most events are insensitive

to the core inlet temperature. The events that

are impacted by lower core inlet temperature

are:

Loss of condenser vacuum (LOCV),

Part length CEA drop,

Single CEA withdrawal within deadband,

and

CEA ejection.

The LOCV event has been reanalyzed for

the upcoming Cycle (Cycle 8) and the results

indicate that the peak RCS pressure remains

below the acceptable limit (110% of the

design pressure, i.e., 2750 psia). The

reactivity anomaly events (remaining events)

will be reanalyzed as part of COLSS/CPC

setpoint calculations. These calculations will

be performed prior to Cycle 8 startup and

will address the impact of the 3°F reduction

on the minimum core inlet temperature. The

CPC/COLSS databases and/or addressable

constants will be modified, as needed due to

proposed change, prior to cycle startup.

A qualitative assessment of the impact of

the proposed change on the calculated LOCA

blowdown loads that are applied to the major

NSSS components, their supports and the

reactor vessel internals was also performed.

This assessment consisted of an evaluation of

the design margins on the major components

and a determination of the impact this lower

temperature would have on those margins.

The evaluation concluded that the impact of

a 3°F cold leg temperature reduction will be

well within the current design margins.

Therefore, the proposed change will not

involve a significant increase in the

probability or consequences of any accident

previously evaluated.

The proposed change to the minimum core

inlet temperature does not involve any

change to any equipment or the manner in

which the plant will be operated. Since no

hardware modifications or changes in

operation procedures will be made, the

proposed change would not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. Therefore, the proposed change

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The impact of the proposed change on the

Waterford 3 FSAR analyses have been

evaluated. The evaluation showed that the

events that were impacted were important

with respect to RCS pressure and fuel

thermal limits. One of the events that was

impacted by the proposed change was the

LOCV event. This event was analyzed and

the results showed that the peak RCS

pressure remained below the acceptable

limit. The impact of this change on other

events (reactivity anomaly events) will be

evaluated as part of the COLSS/CPC setpoint

calculations and the COLSS/CPC databases

and/or addressable constants will modified

as needed to account for any adverse impact

on the results of these events due to the

proposed change.

The impact of this change on the Linear

Heat Generation Rate limits which varies as

a function of the cold leg temperature, is

accounted for by Technical Specification

3.2.1, ‘‘Linear Heat Rate’’. The impact of this

change on LOCA blowdown loads were

evaluated to be insignificant compared to the
Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices 25707

current design margins. Therefore, the

proposed change will not involve a

significant reduction in a margin of safety,

specifically fuel thermal limits and RCS

pressure limit.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122

Attorney for licensee: N.S. Reynolds,

Esq., Winston & Strawn 1400 L Street

N.W., Washington, D.C. 20005-3502

NRC Project Director: William D.

Beckner

Florida Power and Light Company,

Docket Nos. 50-250 and 50-251, Turkey

Point Plant Units 3 and 4, Dade County,

Florida

Dates of amendment request: March

20, 1996, and April 23, 1996

Description of amendment request:

The licensee proposed to change the

Turkey Point Units 3 and 4 Technical

Specifications (TS) to relocate the

requirements for surveillance testing of

the water level and pressure channel

instrumentation for the reactor coolant

system accumulators and clarify the

remaining TS surveillance tests. These

amendments also modify the existing

action statements of TS 3.5.1 for

accumulators to reflect the requirements

of NUREG-1431 by requiring a 72-hour

period to restore boron concentration if

it is not within the limits, and a 1-hour

period to restore any other condition

rendering the accumulators inoperable.

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 proposed amendments do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated because the proposed amendments

conform to the guidance given in Enclosure

1 of the NRC GL [Generic Letter] 93-05. The

overall functional capabilities of the

Emergency Core Cooling System (ECCS)

accumulators will not be modified by the

proposed change. This amendment will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated for the following

reasons:

1) The Water Level and Pressure Channel

Instrumentation does not perform a specific

safety function, and merely provides an

indicating function. The instrumentation in

no way affects the capability of the

accumulators to perform their respective

safety function.

2) The changes in most of the ACTION

statements are more restrictive than current

TS requirements due to the one hour vice

four hour completion time, and therefore will

not increase the probability or consequences

of a previously evaluated accident. If one

accumulator is inoperable for a reason other

than boron concentration, the accumulator

must be returned to OPERABLE status within

1 hour. In this condition, the required

contents of three accumulators cannot be

assumed to reach the core during a Loss Of

Coolant Accident (LOCA). Due to the severity

of the consequences should a LOCA occur in

these conditions, the 1 hour completion time

to open the valve, remove power to the valve,

or restore the proper water volume or

nitrogen cover pressure ensures that prompt

action will be taken to return the inoperable

accumulator to OPERABLE status. The

completion time minimizes the potential for

exposure of the plant to a LOCA under these

conditions. The 1 hour requirement for

restoring a closed isolation valve is merely a

clarification of the existing ‘‘immediate’’ time

requirement.

3) In the case of low-out-of-specification

boron concentration in one accumulator, it

must be returned to within the limits within

72 hours. In this condition, ability to

maintain subcriticality or minimum boron

precipitation time may be reduced. The

boron in the accumulators contributes to the

assumption that the combined ECCS water in

the partially recovered core during the early

reflooding phase of a large break LOCA is

sufficient to keep that portion of the core

subcritical. One accumulator below the

minimum boron concentration limit,

however, will have no effect on available

ECCS water and an insignificant effect on

core subcriticality during reflood. Boiling of

ECCS water in the core during reflood

concentrates boron in the saturated liquid

that remains in the core. In addition, current

Turkey Point analysis demonstrate that the

accumulators discharge only a small amount

following a large main steam line break.

Therefore, their impact on boron

concentration in the reactor coolant system is

minor and not a design limiting event. Thus,

72 hours is allowed to return the boron

concentration to within limits and does not

increase 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 use of the modified specifications can

not create the possibility of a new or different

kind of accident from any previously

evaluated since the proposed amendments

will not change the physical plant or the

modes of plant operation defined in the

facility operating license. No new failure

mode is introduced due to the surveillance

changes and clarifications, since the

proposed changes do not involve the

addition or modification of equipment nor do

they alter the design or operation of affected

plant systems.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The operating limits and functional

capabilities of the affected system are

unchanged by the proposed amendment. The

modified specifications which remove

surveillance requirements from the TS to

plant procedures are consistent with the NRC

GL 93-05 line-item improvement guidance do

not significantly reduce any of the margins of

safety even though the amount of

surveillances is decreased. The modification

of the existing ACTION Statements do not

have an adverse on [sic] affect on the margin

of safety for the following reasons:

1) The SI [Safety Injection] Accumulator

Water Level and Pressure Channel

instrumentation performs no safety function.

2) The changes in ACTION statements a)

and b) are for the most part more restrictive

than existing TS requirements, the reason

being the removal of instrumentation

requirements for operability.

3) In the case of low-out-of-specification

boron concentration in one accumulator, the

requirement will be less restrictive, but the

low boron concentration in one accumulator

will have no effect on available ECCS water

and an insignificant effect on core

subcriticality during reflood and therefore

will not significantly reduce 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 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: Florida International

University, University Park, Miami,

Florida 33199

Attorney for licensee: J. R. Newman,

Esquire, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036

NRC Project Director: Frederick J.

Hebdon

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: April 19,

1996

Description of amendment request:

The proposed amendment would

include revisions to Technical

Specification (TS) 3.3.6.1, ‘‘Primary

Containment and Drywell Isolation

Instrumentation≥; TS 3.3.6.2,

‘‘Secondary Containment Isolation

Instrumentation≥; TS 3.3.7.1, ‘‘Control

Room Ventilation System

Instrumentation≥; TS 3.6.1.2, ‘‘Primary

Containment Air Locks≥; TS 3.6.1.3,
25708 Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices

‘‘Primary Containment Isolation

Valves≥; TS 3.6.4.1, ‘‘Secondary

Containment≥; TS 3.6.4.2, ‘‘Secondary

Containment Isolation Dampers≥; TS

3.6.4.3, ‘‘Standby Gas Treatment≥; TS

3.7.3, ‘‘Control Room Ventilation≥; and

TS 3.7.4, ‘‘Control Room AC System.’’

These TSs would be revised to eliminate

CORE ALTERATIONS as an applicable

condition for which the associated

Limiting Conditions for Operation

(LCO) must be met. Consistent changes

are also proposed for the associated

ACTIONS in each of these LCOs, to

reflect the changes in the applicable

conditions. The intent of these proposed

changes is to allow certain activities

such as control rod venting, which is

considered a CORE ALTERATION in

MODE 5, to be performed without the

requirements of the identified LCOs

being met.

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 eliminate CORE

ALTERATIONS as an applicable condition

requiring operability of the primary and

secondary containment and control room

ventilation system. As stated in the BASES

for the associated Technical Specifications,

operability of these systems is primarily

required for mitigation of the design basis

accident - fuel handling accident (DBA-FHA)

and design basis accident - loss of coolant

accident (DBA-LOCA). The performance of

CORE ALTERATIONS alone is neither a

precursor to, nor a condition during which

these DBAs are postulated to occur. The

proposed changes only delete CORE

ALTERATIONS as an applicable condition

for the affected Technical Specifications. All

other applicable MODES or specified

conditions, including operations with the

potential for draining the reactor vessels

(OPDRVs) and the movement of irradiated

fuel assemblies within the primary or

secondary containment, remain unchanged.

Further, the limitations placed on the

handling of light loads are also unchanged.

The Technical Specifications (and the

separate requirements imposed on the

handling of light loads) will thus continue to

require that systems or functions designed to

mitigate design-basis/previously evaluated

accidents are OPERABLE during the relevant

operating MODES or conditions. On the basis

of the above, it is concluded that the

requested amendment will not increase the

probability or consequences of any accident

previously evaluated.

2. The proposed changes do not involve

any modification to the plant design or to the

operation of plant systems (except to

determine when certain analyzed accidentmitigating systems or features are required to

be OPERABLE). The failure modes

considered for the proposed changes are the

same as those previously considered,

therefore, it can be concluded that no new

failure modes will be created. On this basis,

the proposed amendment will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The changes being made to eliminate

CORE ALTERATIONS as an applicable

condition for which certain LCOs must be

met, do not eliminate the requirements for

operability of those systems or features

assumed to mitigate design-basis or analyzed

accidents during the applicable MODES

when such systems or features are assumed

to be available for performing their mitigating

function. The safety margins assumed or

established by the accident analyses for those

design-basis events (as described in the

accident analyses of the Clinton Power

Station Updated Final Safety Analysis

Report) therefore remain unchanged. Further,

the proposed changes do not impact the

controls imposed on the handling of light

loads (including unirradiated fuel

assemblies) for ensuring that such activities

cannot result in an event that yields

consequences more severe than those

calculated for the DBA-FHA. With respect to

reactivity concerns during refueling

operations (MODE 5), all systems or features

required to be OPERABLE for precluding

inadvertent criticality and monitoring

reactivity changes will continue to be

required OPERABLE as per the current

Technical Specification requirements. The

deletion of CORE ALTERATIONS as an

applicable condition only applies to the

noted systems which do not contribute to

precluding reactivity events. Based on the

above, the proposed changes do not involve

a significant reduction in the margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, Illinois 61727

Attorney for licensee: Leah Manning

Stetener, Vice President, General

Counsel, and Corporate Secretary, 500

South 27th Street, Decatur, Illinois

62525

NRC Project Director: Gail H. Marcus

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: May 1,

1996

Description of amendment request:

The proposed amendment would revise

the Clinton Power Station (CPS)

Operating License and Technical

Specifications (TS) to implement 10

CFR Part 50, Appendix J - Option B, by

referring to Regulatory Guide 1.163,

‘‘Performance-Based Containment LeakTest Program.’’ Specifically, changes

would be made to paragraph 2.D of the

Operating License; TS Section 1.1,

‘‘Definitions;’’ TS 3.6.1.1, ‘‘Primary

Containment;’’ TS 3.6.1.1, ‘‘Primary

Containment Air Locks;’’ TS 3.6.1.3,

‘‘Primary Containment Isolation Valves

(PCIVs);’’ and TS Section 5.5, ‘‘Programs

and Manuals.’’

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 implements new

Option B of 10 CFR 50 Appendix J for

performance-based primary containment

leakage testing. The proposed change does

not involve a change to the plant design or

operation. As a result, the proposed change

does not affect any parameters or conditions

that contribute to the initiation of any

accidents previously evaluated. Thus, the

proposed change cannot increase the

probability of any accident previously

evaluated.

The proposed change potentially affects

the leak-tight integrity of the primary

containment structure which is designed to

mitigate the consequences of a loss-of-coolant

accident (LOCA) by limiting the release of

fission products contained in the post-LOCA

primary containment atmosphere. Functional

integrity of the primary containment must be

maintained during and following the peak

transient pressures and temperatures that

may result from a LOCA. Because the

proposed change does not alter the plant

design, including the primary containment

and primary containment penetrations, and

because it only affects the frequency of

measuring Type A, B, and C leakage without

changing the acceptance criteria for the Type

A, B, and C leakage rate tests, the proposed

change does not directly result in an increase

in the primary containment leakage.

However, decreasing the test frequency can

increase the probability that an increase in

primary containment leakage could go

undetected for an extended period of time.

To minimize that probability, test intervals

will be established based on the performance

history of components being tested.

NUREG-1493, ‘‘Performance-Based

Containment Leak-Test Program,’’ provides

the technical basis for the NRC’s rulemaking

to revise primary containment leakage testing

requirements for nuclear power reactors in 10

CFR 50, Appendix J. NUREG-1493

documents the NRC’s determination that the

effect of primary containment leakage on

overall accident risk is minimal since risk is

dominated by accident sequences that result

in failure of bypass of primary containment.

NUREG-1493 also documents that increasing

the Type A leakage test intervals would have

a minimal impact on public risk, and that

Type B and C tests can identify the vast

majority (greater than ninety five percent) of

all leakage paths. Therefore, performancebased alternatives to current local leakagetesting requirements are feasible without

significant risk impacts.
Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices 25709

Based on the above, IP has concluded that

the proposed change will not result in a

significant increase in the probability or

consequences of any accident previously

evaluated.

2. The proposed change does not involve

a change to the plant design or operation. As

a result, the proposed change does not affect

any of the parameters or conditions that

could contribute to initiation of any

accidents. This change involves the

reduction of Type A, B, and C test frequency.

Except for the method of defining the test

frequency, the methods for performing the

actual tests are not changed. No new accident

modes are created by extending the testing

intervals. No safety-related equipment or

safety functions are altered as a result of this

change. Thus, extending the test frequency

has no influence on, nor does it contribute

to the possibility of a new or different kind

of accident or malfunction from those

previously analyzed.

Based on the above, IP has concluded that

the proposed change will not create the

possibility of a new or different kind of

accident not previously evaluated.

3. The request does not involve a

significant reduction in a margin to safety.

The proposed change only affects the

frequency of the Type A, B, and C testing.

Except for the method of defining the test

frequency, the methods for performing the

actual tests are not changed. However, the

proposed change can increase the probability

that an increase in primary containment

leakage could go undetected for an extended

period of time. NUREG-1493 has determined

that under several different accident

scenarios, the increased risk of radioactivity

release from primary containment is

negligible with the implementation of these

proposed changes.

The margin of safety that has the potential

of being impacted by the proposed change

involves the offsite dose consequences of

postulated accidents which are directly

related to the rate of primary containment

leakage. The primary containment isolation

system is designed to limit leakage to La,

which is defined by the CPS Technical

Specifications to be 0.65% of primary

containment air weight per day at the

calculated peak containment internal

pressure for the design basis loss of coolant

accident (Pa). The limitation on the rate of

primary containment leakage is designed to

ensure that the total leakage volume will not

exceed the value assumed in the accident

analyses at the peak accident pressure (Pa).

The margin of safety for the offsite dose

consequences of postulated accidents

directly related to the primary containment

leakage rate is maintained by continuing to

meet the 1.0 La acceptance criteria. The La

value is not being modified by this proposed

change.

Except for the method of defining the test

frequency, no change in the method of testing

is being proposed. The Type A, B, and C tests

will continue to be done at full pressure (Pa)

or greater. Other programs are in place to

ensure that proper maintenance and repairs

are performed during the service life of the

primary containment and systems and

components penetrating the primary

containment.

As a result, IP has concluded that the

proposed change will not result in a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, Illinois 61727

Attorney for licensee: Leah Manning

Stetener, Vice President, General

Counsel, and Corporate Secretary, 500

South 27th Street, Decatur, Illinois

62525

NRC Project Director: Gail H. Marcus

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: January

25, 1996

Description of amendment request:

The amendment proposes to extend

instrumentation and miscellaneous

surveillance test intervals (STI) to

support 24-month operating cycles.

Additionally, this application proposes:

(1) to revise the Trip Level Settings for

Emergency Bus Loss of Voltage and

Degraded Voltage Instrumentation, (2) to

revise the Reactor Protection System

(RPS) Normal Supply Electrical

Protection Assembly (EPA)

Undervoltage Trip Setpoint, and (3) to

make editorial revisions, clarification

and Bases changes.

Basis for proposed no significant

hazards consideration determination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

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 STI changes evaluated in

Section IV.A do not involve any physical

changes to the plant, do not alter the way

these systems function, and will not degrade

the performance of the plant safety systems.

Proposed instrument setpoint changes ensure

that plant safety limits are not exceeded due

to instrument drift predicted for the longer

calibration interval. The type of testing and

the corrective actions required if the subject

surveillances fail remains the same. The

proposed changes do not adversely affect the

reliability of these systems or affect the

ability of the systems to meet their design

objectives. A historical review of surveillance

test results supports these conclusions.

The Trip Level Setpoint changes evaluated

in Section IV.B ensure that the related

systems perform as assumed in the transient

and accident analysis by ensuring that plant

safety limits are not exceeded due to

instrument drift predicted for the longer

calibration interval. The changes do not alter

the system function, and will not degrade the

performance of plant safety systems. The

proposed Trip Level Setting changes do not

adversely affect the reliability of these

systems or adversely affect the ability of

these systems to meet their design objectives.

The editorial, clarification and Bases

changes evaluated in Section IV.C propose

enhancements that clarify the Technical

Specifications requirements and are editorial

in nature. These changes do not alter any

Technical Specification requirement, do not

involve physical changes to the plant, or alter

any operational setpoints. There are no safety

implications in these proposed changes.

2. create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed STI changes evaluated in

Section IV.A do not modify the design or

operation of the plant, therefore, no new

failure modes are introduced. Proposed

instrument setpoint changes ensure that

plant safety limits are not exceeded due to

instrument drift resulting from the longer

calibration interval. No changes are proposed

to the type and method of testing performed,

only to the length of the surveillance test

interval. Past equipment performance and

on-line testing indicate that longer test

intervals will not degrade these systems. A

historical review of surveillance test results

supports these conclusions.

The Trip Level Setpoint changes evaluated

in Section IV.B ensure that the related

systems perform as assumed in the transient

and accident analysis by ensuring that plant

safety limits are not exceeded due to

instrument drift predicted for the longer

calibration interval. The changes do not alter

the system function, introduce any new

failure modes, and will not degrade the

performance of plant safety systems. The

proposed Trip Level Setting changes do not

adversely affect the reliability of these

systems or adversely affect the ability of

these systems to meet their design objectives.

The editorial, clarification and Bases

changes evaluated in Section IV.C propose

enhancements that clarify the Technical

Specifications requirements and are editorial

in nature. These changes do not alter any

Technical Specification requirement, do not

involve physical changes to the plant, or alter

any operational setpoints. There are no safety

implications in these proposed changes.

3. involve a significant reduction in a

margin of safety.

Although the proposed STI changes

evaluated in Section IV.A will result in an

increase in the interval between surveillance

tests, the impact on system reliability is

minimal. This is based on more frequent online testing and the redundant design of the

evaluated systems. A review of past

surveillance history has shown no evidence
25710 Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices

of failures which would significantly impact

the reliability of these systems. Operation of

the plant remains unchanged by these

proposed STI extensions. The assumptions in

the Plant Licensing Basis are not adversely

impacted. Therefore, the proposed changes

do not result in a significant reduction in the

margin of safety.

The Trip Level Setpoint changes evaluated

in Section IV.B ensure that the related

systems perform as assumed in the transient

and accident analysis by ensuring that plant

safety limits are not exceeded due to

instrument drift predicted for the longer

calibration interval. The changes do not alter

the system function, introduce any new

failure modes, and will not degrade the

performance of plant safety systems. The

proposed Trip Level Setting changes do not

adversely affect the reliability of these

systems or adversely affect the ability of

these systems to meet their design objectives.

The editorial, clarification and Bases

changes evaluated in Section IV.C propose

enhancements that clarify the Technical

Specifications requirements and are editorial

in nature. These changes do not alter any

Technical Specification requirement, do not

involve physical changes to the plant, or alter

any operational setpoints. There are no safety

implications in these proposed changes.

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: Susan Frant

Shankman, Acting

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: April 24,

1996

Description of amendment request:

This amendment proposes to relocate

Technical Specification (TS) 3.11.B/

4.11.B ‘‘Crescent Area Ventilation’’ and

associated Bases from the TS to an

Authority controlled procedure.

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

will not involve a significant hazards

consideration as defined in 10 CFR 50.92,

based on the following:

(1) These changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because:

No modifications, no changes to operating

procedure requirements, and no reduction in

equipment reliability are being made as a

result of these changes. Operating limitations

will continue to be imposed, and required

surveillance will continue to be performed in

accordance with regulations, and written

procedures and instructions that are

auditable by the [Nuclear Regulatory

Commission] NRC. Crescent Area Ventilation

operability and testing requirements will

continue to be an integral part of FitzPatrick

plant operation.

Although future changes to the Crescent

Area Ventilation system will no longer be

controlled by 10 CFR 50.90, proposed

changes will be evaluated under 10 CFR

50.59 and plant procedures. Programmatic

controls will continue to assure that Crescent

Area Ventilation system changes will not

adversely affect [Emergency Core Cooling

System] ECCS or [Reactor Core Isolation

Cooling] RCIC system operability. As such,

there is no significant increase in the

probability or consequences of an accident

previously evaluated.

(2) These changes do not create the

possibility of a new or different type of

accident previously evaluated because:

No modifications, no changes to operating

procedure requirements, and no reduction in

equipment reliability are being made as a

result of these changes. Compliance with

Crescent Area Ventilation system operability

and surveillance requirements will be

assured by maintaining them in an Authority

controlled procedure. Changes to the

Crescent Area Ventilation system will be

subject to the requirements of 10 CFR 50.59.

Therefore, the proposed changes do not

introduce any failure mechanism of a

different type than those previously

evaluated since there are no changes being

made to the facility and do not create the

possibility of a new or different type of

accident previously evaluated.

(3) The proposed amendment does not

involve a reduction in a margin of safety

because:

The Crescent Area Ventilation system

supports Core Spray, [Low Pressure Coolant

Injection] LPCI mode of [Residual Heat

Removal] RHR, containment cooling mode of

RHR, [High Pressure Coolant Injection] HPCI,

and RCIC operability, and Crescent Area

Ventilation system inoperability does affect

these systems. As a result, the requirement

for Crescent Area Ventilation to be operable

for these systems to be considered operable

is implicit in TS Sections 3.5.A, 3.5.B, 3.5.C,

3.5.E, and the definition of OPERABLE

contained in TS Section 1.0.J. 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: Mr. Charles M.

Pratt, 1633 Broadway, New York, New

York 10019.

NRC Project Director: Susan Frant

Shankman, Acting

Public Service Electric & Gas Company,

Docket No. 50-311, Salem Nuclear

Generating Station, Unit No. 2, Salem

County, New Jersey

Date of amendment request: May 7,

1996

Description of amendment request:

The proposed amendment involves a

one-time change to Technical

Specification (TS) 3/4.7.6, ‘‘Control

Room Emergency Air Conditioning

System.’’ The change would permit

refueling of Salem, Unit 2, with the

Control Room Emergency Air

Conditioning System (CREACS)

inoperable in Modes 5 and 6. The

change will expire after the completion

of the Control Room and CREACS

upgrade, which is currently in progress,

and the restart and entry into Mode 4 of

Unit 2 from the current outage.

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 CREACS is not an accident initiator.

CREACS functions post-accident to provide

cooling for Control Room equipment and

habitability for operations personnel.

Therefore, CREACS has no influence on the

probability of any of the previously evaluated

accidents or the other events evaluated as

listed below.

Event

Fuel Handling Accident (Salem)

Waste Gas or Volume Control Tank

Failures

Uncontrolled Boron Dilution

Loss of Offsite Power

Fuel Handling Accident (Hope Creek)

Liquid and Gaseous Waste Releases (Hope

Creek)

Loss of Coolant Accident (LOCA) (Hope

Creek)

Chemical Storage

Barge Collision

Control Room Internal and External Fire

Loss of Spent Fuel Pool Cooling

Loss of Decay Heat Removal

The Control Area Air Conditioning System

(CAACS) and other measures will be
Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices 25711

available to maintain Control Room Envelope

(CRE) ambient temperatures and habitability.

The proposed one-time change does not

impact the consequences of an accident

previously evaluated based on the following

discussions.

The fuel has decayed to such low levels for

more than six months that doses associated

with the fuel handling accident are well

within the limits of GDC [General Design

Criteria] 19. There is insufficient activity

remaining in either gaseous waste storage or

liquid waste storage to force a Control Room

evacuation. In the event of a Loss of Offsite

Power (LOOP), uncontrolled boron dilution

event, loss of spent fuel pool cooling or loss

of decay heat removal, CREACS is not

required in Modes 5 or 6 to mitigate the

consequences of this event and CRE

habitability will be maintained.

For a Hope Creek fuel handling accident,

gaseous radwaste release of LOCA, dose to

Salem Control Room personnel will not

exceed GDC 19 limits. PSE&G [Public Service

Electric & Gas] will maintain the CAACS

[Control Area Air Conditioning System]

outside air intakes either isolated or capable

of being isolated in the event of a Hope Creek

LOCA. The Hope Creek Event Classification

Guide (ECG) requires notification of the

Salem Control Room in the event of an

emergency that has the potential to result in

a radioactive release. The Salem Control

Room will isolate the outside air intakes if

isolation has not already been accomplished.

For the other events evaluated, the need for

evacuation is not considered credible for any

event with the exception of an internal or

external fire. However, the possibility of

evacuation of the CRE in the event of an

internal or external fire would be no different

whether or not CREACS is operating. In the

event of an internal fire, CAACS will remain

in operation to provide purging of the CRE.

For the case of a possible external fire, the

need for evacuation is not considered

credible because of the short duration of the

CREACS outage and improbability of the

factors which are necessary to require an

evacuation of the Control Room (i.e. wind

direction, wind speed, amount of smoke). If

an external fire is detected, operator action

will be taken to isolate the CRE from outside

air while CAACS remains available. In the

unlikely event that the Control Room would

become uninhabitable due to smoke in the

atmosphere, evacuation procedures would be

followed as in the case of the internal fire.

The one chemical storage type event which

might impact the Control Room, rupture of

an ammonium hydroxide tanker, is

precluded by administrative controls such

that no ammonium hydroxide tanker

deliveries will be allowed during the system

upgrade period.

The CAACS will maintain the current

design function and TS Bases requirements

of the CREACS that the ambient air

temperature does not exceed the allowable

temperature for continuous duty rating for

equipment and instrumentation cooled by

the system for the combined CRE. The

CAACS will be maintained functional while

modification to the CREACS is ongoing to

provide cooling during normal operation and

under postulated accident conditions.

Should the temperature in the CRE exceed

allowable levels (85 Degrees F),

administrative controls will be in place to

require restoration of the temperature to

within acceptable levels using CAACS, and

prevent any Core Alteration activities or

positive reactivity changes until the

temperature is restored to acceptable levels.

Therefore, the proposed one-time TS

change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The CREACS is not an accident initiator.

CREACS functions post-accident to provide

cooling for Control Room equipment and

habitability for operations personnel.

Therefore, CREACS inoperability during

Modes 5 and 6 will not result in the creation

of a new or different kind of accident from

any accident previously evaluated. All

pertinent accidents have been assessed and

no other scenarios dealing with fuel

movement, or the need for an operable

CREACS in Mode 5 or 6, have been deemed

credible.

Therefore, the proposed one-time change

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed one-time change does not

significantly reduce the margin of safety as

defined in the Bases for the TS because (1)

there is no credible event as analyzed in

Salem UFSAR [updated final safety analysis

report] Chapter 15 which can cause an

unacceptable environment in the CRE since

the fuel has been decaying for at least six

months, (2) fuel movement inside the Fuel

Handling Building (FHB) is restricted in

accordance with plant TS unless FHB

ventilation is operable, (3) dose to Salem

control room personnel from a potential

Hope Creek fuel handling accident, gaseous

radwaste release or Loss of Coolant Accident

will not exceed GDC 19 limits (4) the one

event which might impact the Control Room,

rupture of an ammonium hydroxide tanker,

is precluded by administrative controls such

that no ammonium hydroxide tanker

deliveries will be allowed during the

CREACS upgrade period, and (5) in the

unlikely event that Control Room evacuation

is required, there is no impact on operator

ability to mitigate the consequences of an

accident in the current plant configuration.

Therefore, the proposed one-time TS

change does not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the 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: Salem Free Public library, 112

West Broadway, Salem, New Jersey

08079

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW, Washington,

DC 20005-3502

NRC Project Director: John F. Stolz

Southern Nuclear Operating Company,

Inc., Docket No. 50-364, Joseph M.

Farley Nuclear Plant, Unit 2, Houston

County, Alabama

Date of amendment request: March

29, 1996

Description of amendment request:

The proposed amendment would revise

Technical Specification 3/4.4.6 ‘‘Steam

Generators’’ and its associated Bases.

Specifically, the steam generator repair

limit would be modified to clarify that

the appropriate method for determining

serviceability for tubes with outside

diameter stress corrosion cracking at the

tube support plate is by a methodology

that more reliably assesses structural

integrity. This amendment request is in

accordance with NRC’s Generic Letter

95-05, ‘‘Voltage-Based Repair Criteria

for Westinghouse Steam Generator

Tubes Affected by Outside Diameter

Stress Corrosion Cracking.’’

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 Farley units in accordance

with the proposed license amendment does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Testing of model boiler specimens for free

standing tubes at room temperature

conditions shows burst pressures as high as

approximately 5000 psi for indications of

outer diameter stress corrosion cracking with

voltage measurements as high as 26.5 volts.

Burst testing performed on pulled tubes,

including tubes pulled from Farley Unit 2,

with up to 7.5 volt indications show burst

pressures in excess of 5300 psi at room

temperature. As stated earlier, tube burst

criteria are inherently satisfied during normal

operating conditions by the presence of the

tube support plate. Furthermore, correcting

for the effects of temperature on material

properties and minimum strength levels (as

the burst testing was done at room

temperature), tube burst capability

significantly exceeds the R.G. [Regulatory

Guide] 1.121 criterion requiring the

maintenance of a margin of 1.43 times the

steam line break pressure differential on tube

burst if through-wall cracks are present

without regard to the presence of the tube

support plate. Considering the existing data

base, this criterion is satisfied with bobbin

coil indications with signal amplitudes over

twice the 2.0 volt voltage-based repair

criteria, regardless of the indicated depth

measurement. This structural limit is based

on a lower 95% confidence level limit of the
25712 Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices

data at operating temperatures. The 2.0 volt

criterion provides a conservative margin of

safety to the structural limit considering

expected growth rates of outside diameter

stress corrosion cracking at Farley. Alternate

crack morphologies can correspond to a

voltage so that a unique crack length is not

defined by a burst pressure to voltage

correlation. However, relative to expected

leakage during normal operating conditions,

no field leakage has been reported from tubes

with indications with a voltage level of under

7.7 volts for a 3/4 inch tube with a 10 volt

correlation to 7/8 inch tubing (as compared

to the 2.0 volt proposed voltage-based tube

repair limit). Thus, the proposed amendment

does not involve a significant increase in the

probability or consequences of an accident.

Relative to the expected leakage during

accident condition loadings, the accidents

that are affected by primary-to-secondary

leakage and steam release to the environment

are Loss of External Electrical Load and/or

Turbine Trip, Loss of All AC Power to

Station Auxiliaries, Major Secondary System

Pipe Failure, Steam Generator Tube Rupture,

Reactor Coolant Pump Locked Rotor, and

Rupture of a Control Rod Drive Mechanism

Housing. Of these, the Major Secondary

System Pipe Failure is the most limiting for

Farley in considering the potential for off-site

doses. The offsite dose analyses for the other

events which model primary-to secondary

leakage and steam releases from the

secondary side to the environment assume

that the secondary side remains intact. The

steam generator tubes are not subjected to a

sustained increase in differential pressure, as

is the case following a steam line break event.

This increase in differential pressure is

responsible for the postulated increase in

leakage and associated offsite doses following

a steam line break event. In addition, the

steam line break event results in a bypass of

containment for steam generator leakage.

Upon implementation of the voltage-based

repair criteria, it must be verified that the

expected distributions of cracking

indications at the tube support plate

intersections are such that primary-tosecondary leakage would result in site

boundary dose within the current licensing

basis. Data indicate that a threshold voltage

of 2.8 volts could result in through-wall

cracks long enough to leak at steam line

break conditions. Application of the

proposed repair criteria requires that the

current distribution of a number of

indications versus voltage be obtained during

the refueling outages. The current voltage is

then combined with the rate of change in

voltage measurement and a voltage

measurement uncertainty to establish an end

of cycle voltage distribution and, thus, leak

rate during steam line break pressure

differential. The leak rate during a steam line

break is further increased by a factor related

to the probability of detection of the flaws.

If it is found that the potential steam line

break leakage for degraded intersections

planned to be left in service coupled with the

reduced allowable specific activity levels

result in radiological consequences outside

the current licensing basis, then additional

tubes will be plugged or repaired to reduce

steam line break leakage potential to within

the acceptance limit. Thus, the consequences

of the most limiting design basis accident are

constrained to present licensing basis limits.

2) The proposed license amendment does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Implementation of the proposed voltagebased tube support plate elevation steam

generator tube repair criteria does not

introduce any significant changes to the plant

design basis. Use of the criteria does not

provide a mechanism that could result in an

accident outside of the region of the tube

support plate elevations. Neither a single or

multiple tube rupture event would be

expected in a steam generator in which the

repair criteria have been applied during all

plant conditions. The bobbin probe signal

amplitude repair criteria are established such

that operational leakage or excessive leakage

during a postulated steam line break

condition is not anticipated. Southern

Nuclear has previously implemented a

maximum leakage limit of 150 gpd per steam

generator. The R.G. 1.121 criterion for

establishing operational leakage limits that

require plant shutdown are based upon leakbefore-break considerations to detect a free

span crack before potential tube rupture. The

150 gpd limit provides for leakage detection

and plant shutdown in the event of the

occurrence of an unexpected single crack

resulting in leakage that is associated with

the longest permissible crack length. R.G.

1.121 acceptance criteria for establishing

operating leakage limits are based on leakbefore-break considerations such that plant

shutdown is initiated if the leakage

associated with the longest permissible crack

is exceeded. The longest permissible crack is

the length that provides a factor of safety of

1.43 against bursting at steam line break

pressure differential. A voltage amplitude of

approximately 9 volts for typical outside

diameter stress corrosion cracking

corresponds to meeting this tube burst

requirement at the 95% prediction interval

on the burst correlation. Alternate crack

morphologies can correspond to a voltage so

that a unique crack length is not defined by

the burst pressure versus voltage correlation.

Consequently, a typical burst pressure versus

through-wall crack length correlation is used

below to define the ‘‘longest permissible

crack’’ for evaluating operating leakage

limits.

The single through-wall crack lengths that

result in tube burst at 1.43 times steam line

break pressure differential and steam line

break conditions are about 0.54 inch and 0.84

inch, respectively. Normal leakage for these

crack lengths would range from about 0.4

gallons per minute to 4.5 gallons per minute,

respectively, while lower 95% confidence

level leak rates would range from about 0.06

gallons per minute to 0.6 gallons per minute,

respectively.

An operating leak rate of 150 gpd per steam

generator has been implemented. This

leakage limit provides for detection of 0.4

inch long cracks at nominal leak rates and 0.6

inch long cracks at the lower 95% confidence

level leak rates. Thus, the 150 gpd limit

provides for plant shutdown prior to

reaching critical crack lengths for steam line

break conditions at leak rates less than a

lower 95% confidence level and for three

times normal operating pressure differential

at less than nominal leak rates.

Considering the above, the implementation

of voltage-based plugging criteria will not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3) The proposed license amendment does

not involve a significant reduction in margin

of safety.

The use of the voltage-based tube support

plate elevation repair criteria is demonstrated

to maintain steam generator tube integrity

commensurate with the requirements of

Generic Letter 95-05 and R.G. 1.121. R.G.

1.121 describes a method acceptable to the

NRC staff for meeting GDC [Generic Design

Criteria] 2, 14, 15, 31, and 32 by reducing the

probability of the consequences of steam

generator tube rupture. This is accomplished

by determining the limiting conditions of

degradation of steam generator tubing, as

established by inservice inspection, for

which tubes with unacceptable cracking

should be removed from service. Upon

implementation of the criteria, even under

the worst case conditions, the occurrence of

outside diameter stress corrosion cracking at

the tube support plate elevations is not

expected to lead to a steam generator tube

rupture event during normal or faulted plant

conditions. The most limiting effect would be

a possible increase in leakage during a steam

line break event. Excessive leakage during a

steam line break event, however, is

precluded by verifying that, once the criteria

are applied, the expected end of cycle

distribution of crack indications at the tube

support plate elevations would result in

minimal, and acceptable primary to

secondary leakage during the event and,

hence, help to demonstrate radiological

conditions are less than an appropriate

fraction of the 10 CFR [Part] 100 guideline.

The margin to burst for the tubes using the

voltage-based repair criteria is comparable to

that currently provided by existing Technical

Specifications.

In addressing the combined effects of

LOCA [loss-of-coolant accident] + SSE [safeshutdown earthquake] on the steam generator

component (as required by GDC 2), it has

been determined that tube collapse may

occur in the steam generators at some plants.

This is the case as the tube support plates

may become deformed as a result of lateral

loads at the wedge supports at the periphery

of the plate due to either the LOCA

rarefaction wave and/or SSE loadings. Then,

the resulting pressure differential on the

deformed tubes may cause some of the tubes

to collapse.

There are two issues associated with steam

generator tube collapse. First, the collapse of

steam generator tubing reduces the RCS

[reactor coolant system] flow area through

the tubes. The reduction in flow area

increases the resistance to flow of steam from

the core during a LOCA which, in turn, may

potentially increase Peak Clad Temperature

(PCT). Second, there is a potential the partial

through-wall cracks in tubes could progress

to through-wall cracks during tube

deformation or collapse or that short through-
Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices 25713

wall indications would leak at significantly

higher leak rates than included in the leak

rate assessments.

Consequently, a detailed leak-before-break

analysis was performed and it was concluded

that the leak-before-break methodology (as

permitted by GDC 4) is applicable to the

Farley reactor coolant system primary loops

and, thus, the probability of breaks in the

primary loop piping is sufficiently low that

they need not be considered in the structural

design basis of the plant. Excluding breaks in

the RCS primary loops, the LOCA loads from

the large branch line breaks were analyzed at

Farley and were found to be of insufficient

magnitude to result in steam generator tube

collapse or significant deformation.

Regardless of whether or not leak-beforebreak is applied to the primary loop piping

at Farley, any flow area reduction is expected

to be minimal (much less than 1%) and PCT

margin is available to account for this

potential effect. Based on analyses’ results,

no tubes near wedge locations are expected

to collapse or deform to the degree that

secondary to primary in-leakage would be

increased over current expected levels. For

all other steam generator tubes, the

possibility of secondary-to-primary leakage

in the event of a LOCA + SSE event is not

significant. In actuality, the amount of

secondary-to-primary leakage in the event of

a LOCA + SSE is expected to be less than that

originally allowed, i.e., 500 gpd per steam

generator. Furthermore, secondary-toprimary in-leakage would be less than

primary-to-secondary leakage for the same

pressure differential since the cracks would

tend to tighten under a secondary-to-primary

pressure differential. Also, the presence of

the tube support plate is expected to reduce

the amount of in-leakage.

Addressing the R.G. 1.83 considerations,

implementation of the tube repair criteria is

supplemented by 100% inspection

requirements at the tube support plate

elevations having outside diameter stress

corrosion cracking indications, reduced

operating leakage limits, eddy current

inspection guidelines to provide consistency

in voltage normalization, and rotating probe

inspection requirements for the larger

indications left in service to characterize the

principle degradation mechanism as outside

diameter stress corrosion cracking.

As noted previously, implementation of

the tube support plate elevation repair

criteria will decrease the number of tubes

that must be taken out of service with tube

plugs or repaired. The installation of steam

generator tube plugs or tube sleeves would

reduce the RCS flow margin, thus

implementation of the voltage-based repair

criteria will maintain the margin of flow that

would otherwise be reduced through

increased tube plugging or sleeving.

Considering the above, it is concluded that

the proposed change does not result in a

significant reduction in margin with respect

to plant safety as defined in the Final Safety

Analysis Report or any bases of the plant

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: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302

Attorney for licensee: M. Stanford

Blanton, Esq., Balch and Bingham, Post

Office Box 306, 1710 Sixth Avenue

North, Birmingham, Alabama 35201

NRC Project Director: Herbert N.

Berkow

Southern Nuclear Operating Company,

Inc., Docket No. 50-364, Joseph M.

Farley Nuclear Plant, Unit 2, Houston

County, Alabama

Date of amendment request: April 22,

1996

Description of amendment request:

The proposed amendment would

implement a new F* criterion based on

maintaining existing safety margins for

steam generator tube structural integrity

concurrent with allowance for NDE

(nondestructive examination) eddy

current uncertainty.

Basis for proposed no significant

hazards consideration determination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated. The proposed change retains the

existing margin in the F* distance used to

meet regulatory guidance of draft Regulatory

Guide 1.121 and only changes the amount of

assumed NDE eddy current uncertainty based

on the type of eddy current technology

utilized in the inspection. Therefore, there is

no significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

WCAP 11306, Revision 2, ‘‘Tubesheet Region

Plugging Criterion for the Alabama Power

Company Farley Nuclear Station Unit 2

Steam Generators,’’ provides adequate basis

for the F* distance proposed of 1.54 plus

allowance for eddy current uncertainty

measurement. Since the value of 1.54 inches

was used in the analysis no new or different

kind of accident from any accident

previously evaluated will be created.

3. The proposed change does not involve

a significant reduction in a margin safety.

Since the value of 1.54 inches already is used

in the steam generator tube pull out analysis,

there is no significant change to a margin

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: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302

Attorney for licensee: M. Stanford

Blanton, Esq., Balch and Bingham, Post

Office Box 306, 1710 Sixth Avenue

North, Birmingham, Alabama 35201

NRC Project Director: Herbert N.

Berkow

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, MissouriDate of

application request: February 23, 1996,

as supplemented by letter dated April

24, 1996.

Description of amendment request:

The amendment would add a footnote

in the license for Callaway Plant, Unit

No. 1 to indicate that Union Electric

Company has entered into a merger

agreement with CIPSCO Incorporated

which provides for Union Electric

Company to become a wholly-owned

operating company of Ameren

Corporation, a registered public utility

holding company under the Public

Utility Holding Company Act of 1935,

as amended. After the merger, Union

Electric Company would continue to

own and operate the Callaway Plant as

an operating company subsidiary of

Ameren Corporation.

Basis for proposed no significant

hazards consideration determination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change does not affect

accident initiators or assumptions. The

radiological consequences of any accident

previously evaluated remain unchanged. The

change is an administrative change to reflect

Union Electric’s status as an operating

company subsidiary of Ameren.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change does not reduce the

margin of safety assumed in any accident

analysis or affect any safety limits. The

change is administrative and reflects Union

Electric’s status as an operating company

subsidiary of Ameren.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change does not reduce the

margin of safety assumed in any accident
25714 Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices

analysis or affect any safety limits. The

change is administrative and reflects Union

Electric’s status as an operating company

subsidiary of Ameren.

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: 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, D.C. 20037

NRC Project Director: William H.

Bateman

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request: April 30,

1996

Description of amendment request:

The proposed amendment would revise

Kewaunee Nuclear Power Plant (KNPP)

Technical Specification (TS) 3.1.b.1, its

associated bases, and Figure TS 3.1-4 by

extending the low temperature

overpressure protection (LTOP)

requirements through the end of

operating cycle 33 or 33.41 effective full

power years. The only technical change

being proposed is the substitution of

end of life fluence for the end of

operating cycle 21 fluence.

Basis for proposed no significant

hazards consideration determination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The proposed change was reviewed in

accordance with the provisions of 10 CFR

50.92 to show no significant hazards exist.

The proposed change will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The LTOP setpoint and revised P/T

[pressure/temperature] limits reflected in

proposed Figure TS 3.1-4 ensure that the

Appendix G pressure/temperature limits are

not exceeded, and therefore, help ensure that

RCS integrity is maintained. The changes do

not modify the reactor coolant system

pressure boundary, nor make any physical

changes to the facility design, material,

construction standards, or setpoints. The

LTOP valve setpoint remains set at 500 psi.

The LTOP enabling temperature based on

Figure TS 3.1-2 is 338°F and is more

conservative than a value of 303° Figure TS

3.1-4. The LTOP enabling temperature based

on Figure TS 3.1-2 remains unchanged by

this PA [proposed amendment]. The

probability of a LTOP event occurring is

independent of the pressure-temperature

limits for the RCS pressure boundary.

Therefore, the probability of a LTOP event

occurring remains unchanged.

The calculation of pressure temperature

limits in accordance with approved

regulatory methods provides assurance that

reactor pressure vessel fracture toughness

requirements are met and the integrity of the

RCS [reactor coolant system] pressure

boundary is maintained. Similar

methodology was used in calculations to

support approved amendment 120 to the

Kewaunee Technical Specifications dated

April 26, 1995. The material property basis,

including chemistry factor and initial

reference temperature for the unirradiated

material (RTNDT), used for this PA is the same

as that used in the current TS. The only

technical change being made in this PA is the

use of end of life fluence.

The use of predicted fluence values

through the end of operating cycle 33 is

appropriately considered within the

calculations in accordance with standard

industry methodology previously docketed

under WCAP 13227 and WCAP 14279. The

neutron exposure projections utilized for

calculation of the reference temperature were

multiplied by a factor of 1.11 to adjust for

biases observed between cycle specific

calculations and the results of neutron

dosimetry for the four surveillance capsules

removed from the KNPP reactor. The factor

of 1.11 was derived by taking the average of

the measured to calculation (M/C) flux ratios

obtained from the dosimetry results of

capsules V, R, P, and S removed from the

KNPP reactor vessel. The resulting effect of

using predicted fluence values through the

end of cycle 33 instead of cycle 21 is to

require the plant to evaluate LTOP transients

to more limiting requirements. The proposed

PT limits are shifted to a lower pressure and

higher temperature, which is more

conservative.

The changes do not adversely affect the

integrity of the RCS such that its function in

the control of radiological consequences is

affected. In addition, the changes do not

affect any fission barrier. The changes do not

degrade or prevent the response of the LTOP

relief valve or other safety related system to

accidents described in Chapter 14 of the

USAR. In addition, the changes do not alter

any assumption previously made in the

radiological consequences evaluations nor

affect the mitigation of the radiological

consequences of an accident described in the

USAR. Therefore, the consequences of an

accident previously evaluated in the USAR

will not be increased.

Thus, the operation of KNPP Unit 1 in

accordance with the PA does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

2. Create the possibility of a new or

different type of accident from an accident

previously evaluated.

The Appendix G pressure temperature

limitations were prepared using methods

derived from the ASME Boiler and Pressure

Vessel Code and the criteria set forth in NRC

Regulatory Standard Review Plan 5.3.2. The

changes do not cause the initiation of any

accident nor create any new credible limiting

failure for safety-related systems and

components. The changes do not result in

any event previously deemed incredible

being made credible. As such, it does not

create the possibility of an accident different

than any evaluated in the USAR.

The changes do not have any effect on the

ability of the safety-related systems to

perform their intended safety functions. The

changes do not create failure modes that

could adversely impact safety-related

equipment. Therefore, it will not create the

possibility of a malfunction of equipment

important to safety different than previously

evaluated in the USAR. Thus, the PA does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The use of Paragraph (c)(2)(ii)(A) of 10 CFR

50.61, initial reference temperature of -50°F,

and the fluence values through EOC [end of

cycle] 33 does not modify the reactor coolant

system pressure boundary, nor make any

physical changes to the LTOP setpoint or

system design. Proposed Figure TS 3.1-4 was

prepared in accordance with regulatory

requirements and requires evaluation of

LTOP events to more limiting requirements

of neutron exposure projections of 33.41

EFPY instead of 18.40 EFPY.

Therefore, the PA does not create the

possibility of a new or different type of

accident from any accident previously

evaluated.

3. Involve a significant reduction in the

margin of safety.

The Appendix G pressure temperature

limitations were prepared using methods

derived from the ASME Boiler and Pressure

Vessel Code and the criteria set forth in NRC

Regulatory Standard Review Plan 5.3.2.

These documents along with the

calculational limitations specified in 10 CFR

50.61 are an acceptable method for

implementing the requirements of 10 CFR 50

Appendices G and H. Inherent conservatism

in the P/T limits resulting from these

documents include:

a. An assumed defect in the reactor vessel

wall with a depth equal to 1/4 of the

thickness of the vessel wall (1/4T) and a

length equal to 1-1/2 times the thickness of

the vessel wall.

b. Assumed reference flaw oriented in both

longitudinal and circumferential directions

and limiting material property. At KNPP, the

only weld in the core region is oriented in

the circumferential direction.

c. A factor of safety of 2 is applied to the

membrane stress intensity factor.

d. The limiting toughness is based upon a

reference value (KIR) which is a lower bound

on the dynamic crack initiation or arrest

toughness.

e. A 2-sigma margin term is applied in

determining the adjusted reference

temperature (ART) that is used to calculate

the limiting toughness.

Similar methodology was used in

calculations to support approved amendment

120 dated April 26, 1995. Beyond the

conservatism described above, WPSC
Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices 25715

[Wisconsin Public Service Corporation] has

incorporated the following additional margin

in preparing this PA:

a. The neutron exposure projections were

multiplied by a factor of 1.11 to adjust for

biases observed between cycle specific

calculations and the results of neutron

dosimetry for the four surveillance capsules

removed from the KNPP reactor. The factor

of 1.11 was derived by taking the average of

the measured to calculation (M/C) flux ratios

obtained from the dosimetry results of

capsules V, R, P, and S removed from the

KNPP reactor vessel.

b. The calculated material-specific

chemistry factor value is 191.27 and is based

on KNPP surveillance capsule data from

capsules V, R, and P. Utilization of KNPP’s

most recent surveillance capsule data from

capsule S results in chemistry factor value of

190.6. Consistent with calculation C10689,

Revision 1 the value used for chemistry

factor in this PA remains 191.27, which is

conservative.

c. The LTOP enabling temperature based

on Figure TS 3.1-2 is 338°F and is more

conservative than a value of 303°F which is

supported by proposed Figure TS 3.1-4. The

LTOP enabling temperature based on Figure

TS 3.1-2 remains unchanged by this PA.

d. The reactor coolant pump starting

restrictions of TS 3.1.a.1.c remain in place.

An alternative methodology to the safety

margins required by Appendix G to 10 CFR

Part 50 has been developed by the ASME

Working Group on Operating Plant Criteria.

This methodology is contained in ASME

Code Case N-514. The Code Case N-514

provides criteria to determine pressure limits

during LTOP events that avoid certain

unnecessary operational restrictions, provide

adequate margins against failure of the

reactor pressure vessel, and reduce the

potential for unnecessary activation of the

relief valve used for LTOP. Specifically, the

ASME Code Case N-514 allows

determination of the setpoint for LTOP

events such that the maximum pressure in

the vessel would not exceed 110% of the P/

T limits of the existing ASME Appendix G;

and redefines the enabling temperature as a

coolant temperature less than 200°F or a

reactor vessel metal temperature less than

RTNDT + 50°F greater. Code Case N-514,

‘‘Low Temperature Overpressure Protection,’’

has been approved by the ASME Code

Committee but not yet approved for use in

Regulatory Guide 1.147. The content of this

code case has been incorporated into

Appendix G of Section XI of the ASME Code

and published in the 1993 Addenda to

Section XI. It is expected that when the NRC

revises 10 CFR 50.55a, it will endorse the

1993 Addenda and Appendix G of Section XI

into the regulations. As stated above, this PA

utilizes Appendix G limits and an enabling

temperature corresponding to a reactor vessel

metal temperature less than RTNDT + 90°F,

which is more conservative than the

alternative methodology contained in Code

Case N-514.

The revised calculations meet the NRC

acceptance criteria for the LTOP setpoint and

system design as described in NRC Safety

Evaluation Report (SER) dated September 6,

1995 which concluded that ‘‘the spectrum of

postulated pressure transients would be

mitigated...such that the temperature

pressure limits of Appendix G to 10 CFR 50

are maintained.’’

Utilization of methodology set forth in the

ASME Boiler and Pressure Vessel Code, NRC

Regulatory Standard Review Plan 5.3.2, 10

CFR 50.61, and 10 CFR 50 Appendices G and

H with the above additional margins ensures

that proper limits and safety factors are

maintained. Thus, the PA does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, Wisconsin 54311-7001

Attorney for licensee: Bradley D.

Jackson, Esq., Foley and Lardner, P. O.

Box 1497, Madison, Wisconsin 53701-

1497

NRC Project Director: Gail H. Marcus

Wisconsin Public Service Corporation,

Docket No. 50-305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request: May 1,

1996

Description of amendment request:

The proposed amendment would revise

Kewaunee Nuclear Power Plant (KNPP)

Technical Specification (TS) 4.2.b,

‘‘Steam Generator Tubes,’’ its associated

bases, and Figure TS 4.2-1 by redefining

the pressure boundary for Westinghouse

mechanical hybrid expansion joint (HEJ)

steam generator (SG) tube sleeves. The

proposed amendment supersedes in its

entirety a previously submitted

proposed amendment dated October 6,

1995, which was published in the

Federal Register on November 8, 1995

(60 FR 56372).

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:

This proposed change was reviewed in

accordance with the provisions of 10 CFR

50.92 to show no significant hazards exist.

1. Operation of the KNPP in accordance

with the proposed license amendment does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Mechanical testing shows inherent

structural integrity of the HEJ [hybrid

expansion joint] upper joint such that the

tube rupture capability recommendations of

RG [Regulatory Guide] 1.121 are met, even

for instances of 100-percent throughwall, 360

degree degradation in the HRLT [hardroll

lower transition] region. Structural test

results are documented in WCAPs-14157,

-14157 Addendum 1, -14446 and -14641.

Based on this test data, the structural

recommendations of RG 1.121 are satisfied

when there is a difference of at least 0.003

inch, between the maximum hardroll

diameter of the sleeve, and the diameter at

the elevation of the PTI [parent tube

indication] center line; i.e. there is an

interference lip of 0.003 inch or more. The

proposed pressure boundary will allow PTIs

located such that there is a minimum

diameter change of 0.003 inch (not including

an allowance for measurement uncertainty)

between the maximum point of the sleeve

hardroll, and the diameter at the elevation of

the PTI peak amplitude to remain in service.

Based on the high degree of structural

integrity of the HEJ upper joint, it can be

concluded that application of the revised

pressure boundary criteria will not result in

an increased probability of an accident

previously evaluated.

Each sleeved tube with a PTI located in the

HRLT such that there is a change in diameter

of 0.003 inch to 0.013 inch, will be assigned

a conservatively bounding primary-tosecondary SLB [steam line break] leakage

value of 0.025 gpm per indication.

Indications located such that there is a

change in diameter of greater than 0.013 inch

will not contribute to the SLB leakage. The

total number of indications remaining in

service will be limited such that the primaryto-secondary leakage during a postulated SLB

will not exceed a small fraction of the 10 CFR

Part 100 guidelines. For KNPP this has been

calculated to be 34.0 gpm for the faulted

loop. Therefore, it can be concluded that

application of the revised pressure boundary

criteria will not increase the consequences of

an accident previously evaluated.

2. The proposed license amendment

request does not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

Implementation of the revised pressure

boundary will not introduce a change to the

design basis or operation of the plant.

Mechanical testing of degraded sleeve joints

supports the conclusions that the joint

retains structural integrity (tube burst)

capability consistent with RG 1.121, and

leakage integrity with regards to a small

fraction of the 10 CFR Part 100 guidelines. As

with the initial installation of the sleeves,

implementation of the relocated pressure

boundary does not interact with other

portions of the reactor coolant system. Any

hypothetical accident as a result of potential

PTIs is bounded by the existing tube rupture

accident analysis. Neither the sleeve design

nor implementation of the redefined pressure

boundary affects any other component or

location of the tube outside of the immediate

area repaired. Therefore application of the

revised pressure boundary criteria will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed license amendment does

not involve a significant reduction in the

margin of safety.
25716 Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices

The safety factors used in establishment of

the HEJ sleeved tube pressure boundary are

consistent with the safety factors in the

ASME Boiler and Pressure Vessel Code used

in SG [steam generator] design. Based on the

sleeve-to-tube geometry, it is unrealistic to

consider that application of the revised

pressure boundary could result in single tube

leak rates exceeding the normal makeup

capacity during normal operating conditions.

The pressure boundary developed in WCAPs14446 and -14641 have been developed using

the methodology of RG 1.121. The

performance characteristics of the postulated

degraded parent tubes of HEJ sleeve/tube

joints have been verified by testing to retain

structural integrity and preclude significant

leakage during normal and postulated

accident conditions. Testing indicates that

postulated circumferentially separated tubes

which the pressure boundary [addresses]

would not experience axial displacement

during either normal operation or SLB

conditions. The existing offsite dose

evaluation performed for KNPP in support of

the voltage based repair criteria for axial

ODSCC [outside diameter stress corrosion

cracking] at TSP [tube support plate]

intersections established a faulted loop

primary to secondary leak rate of 34.0 gpm.

Following implementation of the criteria,

postulated leakage from all sources must not

exceed 34.0 gpm in the faulted loop.

Maintenance of this limit will ensure that

offsite doses would not exceed the currently

accepted limit of a small fraction of the 10

CFR Part 100 guidelines. The pressure

boundary definition uses a conservatively

established ‘‘per indication’’ leak rate for

estimation of SLB leakage. This leak rate is

applied to all indications left in service

within the HRLT, regardless of indications

length and throughwall extent. Application

of the revised pressure boundary criteria will

not result in a significant reduction in the

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, Wisconsin 54311-7001.

Attorney for licensee: Bradley D.

Jackson, Esq., Foley and Lardner, P. O.

Box 1497, Madison, Wisconsin 53701-

1497

NRC Project Director: Gail H. Marcus

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: July 29,

1994, as superseded by letter dated

September 15, 1995, and supplements

dated March 8, 1996, and April 18, 1996

Description of amendment request:

The proposed amendment revises TS 3/

4.8.1 and its associated Bases to

improve overall emergency diesel

generator reliability and availability.

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.

These proposed changes do not involve a

change in the operational limits or physical

design of the emergency power system.

Emergency diesel generator operability and

reliability will continue to be assured while

minimizing the number of required

emergency diesel generator starts. Also,

emergency diesel generator reliability will be

enhanced by minimizing severe test

conditions which can lead to premature

failures.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

These proposed changes do not involve a

change in the operational limits or physical

design of the emergency power system. The

performance capability of the emergency

diesel generator will not be affected.

Emergency diesel generator reliability and

availability will be improved by the

implementation of the proposed changes.

There is no actual impact on any accident

analysis.

3. The proposed change does not involve

a significant reduction in a margin of safety.

These proposed change do not involve a

change in the operational limits or physical

design of the emergency power system. The

performance capability of the emergency

diesel generator will not be affected.

Emergency diesel generator reliability and

availability will be improved by the

implementation of the proposed changes. No

margin of safety is reduced.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

2300 N Street, N.W., Washington, D.C.

20037

NRC Project Director: William H.

Bateman

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: May 1,

1996

Description of amendment request:

This license amendment request

proposes to revise Section 6.0 of the

technical specifications to reflect

position title changes within the Wolf

Creek Nuclear Operating Corporation

(WCNOC) organization.

Basis for proposed no significant

hazards consideration determination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change does not involve a

significant increase in the probability of

consequences of an accident previously

evaluated. These changes involve

administrative changes to the WCNOC

organization and to the position qualification

of plant personnel.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. This change is administrative in

nature and does not involve a change to the

installed plant systems or the overall

operating philosophy of Wolf Creek

Generating Station.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change does not involve a

significant reduction in a margin of safety.

This change does not involve any changes in

overall organizational commitments. A

position title change alone does not reduce

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

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

2300 N Street, N.W., Washington, D.C.

20037

NRC Project Director: William H.

Bateman
Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices 25717

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.

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of amendment request: April 25,

1996

Brief description of amendment

request: The amendment relocates the

technical specification (TS) Traversing

In-Core Probe System Limiting

Condition for Operation 3/4.3.7.7 and

its Bases 3/4.3.7.7 to the Technical

Requirements Manual, and modifies

Note (f) of TS Table 4.3.1.1-1.

Date of publication of individual

notice in Federal Register: May 8, 1996

(61 FR 20840)

Expiration date of individual notice:

June 7, 1996

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, OES Nuclear, Inc.,

Pennsylvania Power Company, Toledo

Edison Company, Docket No. 50-440,

Perry Nuclear Power Plant, Unit No. 1,

Lake County, Ohio

Date of application for amendment:

April 26, 1996

Brief description of amendment

request: The proposed amendment

would correct minor technical and

administrative errors in the Improved

Technical Specifications prior to its

implementation.

Date of individual notice in Federal

Register: May 9, 1996 (61 FR 21213)

Expiration date of individual notice:

June 10, 1996

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, Ohio

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:

February 1, 1996

Brief description of amendments:

These amendments revised (1)

Technical Specifications (TS) 3/4.1.1.1,

6.9.1.9, and 6.9.1.10 to relocate the

shutdown margin (reactor trip breakers

open) to the Core Operating Limits

Report; (2) TS 3/4.3.2 (Tables 3.3-3 and

3.3-4) to specify an additional

restriction for the allowed lowpressurizer-pressure trip setpoint when

reducing reactor coolant (RCS) system

pressure in Mode 3; (3) TS Section 2.2.1

(Table 2.2-1) to make it consistent with

the footnote in TS Tables 3.3-3 and 3.3-

4; and (4) TS Sections 3/4.5.2 and 3/

4.5.3 to require two emergency core

cooling system subsystems to be

operable in Mode 3 whenever the RCS

cold-leg temperature is equal to or above

485°F. The Table of Contents and the

Bases are also revised to reflect these

changes.

Date of issuance: April 30, 1996

Effective date: April 30, 1996, to be

implemented within 45 days of issuance

Amendment Nos.: Unit 1 - 106; Unit

2 - 98; Unit 3 - 78

Facility Operating License Nos. NPF41, NPF-51, and NPF-74: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 27, 1996 (61 FR 13522)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated April 30, 1996.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona

85004

Duquesne Light Company, et al., Docket

Nos. 50-334 and 50-412, Beaver Valley

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of application for amendments:

December 27, 1995

Brief description of amendments:

These amendments modify Tables 3.3-

11 and 4.3-7 of Beaver Valley Power

Station, Unit Nos. 1 and 2 (BVPS-1 and

BVPS-2) Technical Specification 3.3.3.8

(Accident Monitoring Instrumentation)

such that only one valve position

indication system for the poweroperated relief valves and safety valves

is required to be operable. Minor

editorial changes to BVPS-1 TS 3.3.3.8

and its associated Action Statements are

also being made. These changes make

the requirements of TS 3.3.3.8

consistent with the NRC’s Improved

Standard Technical Specifications

(NUREG-1431, Revision 1) and with the

guidance of Regulatory Guide 1.97,

NUREG-0578, and NUREG-0737.

Date of issuance: May 1, 1996

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment Nos.: 199 and 81
25718 Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices

Facility Operating License Nos. DPR66 and NPF-73: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: January 31, 1996 (61 FR 3499)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 1, 1996No

significant hazards consideration

comments received: No.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 15001.

Duquesne Light Company, et al., Docket

No. 50-334, Beaver Valley Power

Station, Unit No. 1, Shippingport,

Pennsylvania

Date of application for amendment:

February 12, 1996

Brief description of amendment: The

amendment revises Technical

Specification (TS) 4.6.2.2.d to delete the

reference to the specific test acceptance

criteria for the Containment

Recirculation Spray Pumps and replaces

the specific test acceptance criteria with

reference to the requirements of the

Inservice Testing (IST) Program. In

addition, the 18-month test frequency is

replaced with the test frequency

requirements specified in the IST

Program. The amendment also revises

the Bases for TS 4.6.2.2.d to describe

this revision to TS 4.6.2.2.d.

Date of issuance: May 7, 1996

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No: 200

Facility Operating License No. DPR66. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 13, 1996 (61 FR 10393)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 7, 1996 No

significant hazards consideration

comments received: No

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001

Florida Power Corporation, et al.,

Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit No. 3,

Citrus County, Florida

Date of application for amendments:

March 21, 1996 as supplemented April

8, 15, and 18, 1996.

Description of amendment request:

The proposed amendment provides for

interim repair criteria for volumetric

intergranular attack (IGA) indications in

the once-through-steam generators

(OTSG). The interim repair criteria is

based on bobbin coil voltage response

and motorized rotating pancake coil

probe dimensional measurements. The

amendment would be applicable for

IGA indications within the region below

the first tube support plate and the

secondary face of the lower tubesheet

(first span) of the OTSG and for one

cycle only until Refuel 11.

Date of issuance: April 30, 1996

Effective date: April 30,

1996Amendment Nos. 154

Facility Operating License No. DPR72: Amendment revised the Technical

Specifications. Public comments

requested as to proposed no significant

hazards consideration: Yes (61 FR

13888). That notice provided an

opportunity to submit comments on the

Commission’s proposed no significant

hazards consideration determination.

No comments have been received. The

notice also provided for an opportunity

to request a hearing by April 29, 1996,

but indicated that if the Commission

makes a final no significant hazards

consideration determination any such

hearing would take place after issuance

of amendment. The Commission’s

related evaluation of this amendment is

contained in a Safety Evaluation dated

April 30, 1996

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida

32629

Indiana Michigan Power Company,

Docket No. 50-316, Donald C. Cook,

Nuclear Plant, Unit No. 2, Berrien

County, Michigan

Date of application for amendment:

March 12, 1996 (AEP:NRC:1248)

Brief description of amendment: The

amendment removes the technical

specifications related to shutdown and

control rod position indication while in

shutdown modes 3, 4, and 5.

Date of issuance: May 2, 1996

Effective date: May 2, 1996, with full

implementation within 45 days

Amendment No.: 194

Facility Operating License No. DPR74. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: March 27, 1996 (61 FR 13527)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 2, 1996.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085

Nebraska Public Power District, Docket

No. 50-298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request: May 5,

1995 and July 14, 1995, supplemented

by letter dated March 5, 1996

Brief description of amendment: The

amendment revised the Technical

Specifications to 1) verify that the

redundant diesel generator is operable

upon the loss of one diesel generator,

and implement provisions to verify that

the operable diesel generator does not

have a common cause failure; 2)

incorporate provisions to allow a

modified start for the diesel generators;

and 3) remove the requirement that the

reactor power level be reduced to 25%

of rated power upon loss of both diesel

generator units or both incoming power

sources (start-up and emergency

transformers). In addition, the period of

time allowed for continued reactor

operation with both diesels inoperable

was reduced from 24 to two hours.

Date of issuance: April 29, 1996

Effective date: April 29, 1996

Amendment No.: 175

Facility Operating License No. DPR46: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 27, 1995 (60 FR

49939) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

April 29, 1996.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Auburn Memorial Library,

1810 Courthouse Avenue, Auburn, NE

68305.

North Atlantic Energy Service

Corporation, Docket No. 50-443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request:

September 22, 1995

Description of amendment request:

The amendment changes the ACTION

specified in Table 3.3-3, Engineered

Safety Features Actuation System

Instrumentation, from ACTION 18 to

ACTION 15 for Functional Unit 8.b,

Automatic Switchover to Containment

Sump - RWST Level Low-Low.

Date of issuance: May 7, 1996,

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 47

Facility Operating License No. NPF86. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 6, 1995 (60 FR

62493) The Commission’s related
Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices 25719

evaluation of the amendment is

contained in a Safety Evaluation dated

May 7, 1996. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

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:

May 26, 1995, as supplemented October

20, 1995, and May 3, 1996.

Brief description of amendment: The

amendment modifies Technical

Specification (TS) 3.8.1.2, ‘‘Electrical

Power Systems, Shutdown,’’ TS 3.8.2.2,

‘‘Electrical Power Systems, A.C.

Distribution - Shutdown,’’ and TS

3.8.2.4, ‘‘Electrical Power Systems, D.C.

Distribution - Shutdown,’’ to provide

operational flexibility as well as

consistency between action statements

and to eliminate certain surveillance

requirements that are not applicable in

Mode 5 or 6.

The proposed changes relating to TS

3.8.1.1, ‘‘Electrical Power Systems, A.C.

Sources, Operating,’’ are not included in

this amendment since this portion of the

TS change is still under review by the

staff and will be addressed at a later

date.

Date of issuance: May 6, 1996

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 197

Facility Operating License No. DPR65. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 6, 1995 (60 FR

62493) The October 20, 1995, letter

formally withdrew the need for exigent

handling of the May 26, 1995, request

and requested an additional change to

TS 3.8.2.4. The May 3, 1996, letter

withdrew a portion of the initial request

which did not affect the initial proposed

no significant hazards consideration.

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 6, 1996. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360, and Waterford

Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, CT 06385.

Power Authority of the State of New

York, Docket No. 50-286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of application for amendment:

March 14, 1996

Brief description of amendment: The

amendment allows a one-time extension

of the intervals for the pressurizer safety

valve setpoint and snubber functional

testing that is due in May 1996.

Date of issuance: May 3, 1996

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 165

Facility Operating License No. DPR26: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 3, 1996, (61 FR 14835)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 3, 1996.No

significant hazards consideration

comments received: No

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

Public Service Electric & Gas Company,

Docket Nos. 50-272 and 50-311, Salem

Nuclear Generating Station, Unit Nos. 1

and 2, Salem County, New Jersey

Date of application for amendments:

January 4, 1996

Brief description of amendments: The

amendments change Technical

Specification 3/4.8.2.5, ‘‘28-Volt D.C.

Distribution - Operating.’’ The

amendment for Unit 1 makes Unit 1

requirements similar to Unit 2 by

defining the specific battery chargers

that are required for each train and by

restricting the use of the backup battery

charger to 7 days. The amendments for

both units also require that the 28-Volt

DC bus be energized for that bus to be

OPERABLE.

Date of issuance: April 29, 1996

Effective date: Both units, as of date

of issuance, to be implemented within

60 days.Amendment Nos. 182 and 163

Facility Operating License Nos. DPR70 and DPR-75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: February 14, 1996 (61 FR

5818) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

April 29, 1996No significant hazards

consideration comments received: No

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, New Jersey

08079

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

September 6, 1995, as supplemented by

letters dated January 30, 1996, March

27, 1996, and April 2, 1996.

Brief description of amendment: The

amendment revises TS 5.3.1 to reflect a

change in the maximum initial

enrichment for reload fuel, subject to

the integral fuel burnable absorber

(IFBA) requirements, and a change in

the maximum fuel enrichment not

requiring IFBAs. The amendment also

changes the maximum reference kinfinity

in TS 5.6.1.1 for fuel storage in Region

1 of the spent fuel pool and revises TS

Figure 3.9-1 to reflect a change to the

maximum initial enrichment for fuel

stored in Region 2 of the spent fuel pool.

Date of issuance: April 30, 1996

Effective date: April 30, 1996, to be

implemented within 30 days from the

date of issuance.

Amendment No.: 109

Facility Operating License No. NPF30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: November 8, 1995 (60 FR

56372). The January 30, 1996, March 27,

1996, and April 2, 1996, supplemental

letters provided additional clarifying

information and did not change the

original no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated April 30, 1996.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

February 9, 1996

Brief description of amendment: The

amendment revised Technical

Specification 5.3.1 to allow the use of

ZIRLO clad fuel rods and ZIRLO filler

rods.

Date of issuance: April 30, 1996

Effective date: April 30, 1996, to be

implemented within 30 days of

issuance.

Amendment No.: 110

Facility Operating License No. NPF30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: February 28, 1996 (61 FR

7558) The Commission’s related
25720 Federal Register / Vol. 61, No. 100 / Wednesday, May 22, 1996 / Notices

evaluation of the amendment is

contained in a Safety Evaluation dated

April 30, 1996.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Virginia Electric and Power Company,

Docket Nos. 50-280 and 50-281, Surry

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia.

Date of application for amendments:

January 30, 1996

Brief description of amendments:

These amendments modify the

Technical Specifications requirements

for the sampling of the reactor coolant

for dissolved oxygen chlorides and

fluorides.

Date of issuance: 209 and 209

Effective date: April 29, 1996

Amendment Nos. 209 and 209

Facility Operating License Nos. DPR32 and DPR-37: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: March 27, 1996 (61 FR 13533)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated April 29, 1996.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185

Washington Public Power Supply

System, Docket No. 50-397, Nuclear

Project No. 2, Benton County,

Washington

Date of application for amendment:

January 19, 1996, as supplemented by

letter dated March 19, 1996.

Brief description of amendment: The

amendment modifies the Technical

Specifications for leak tests of

containment isolation valves. The

amendment replaces the current

specified surveillance intervals for

containment leak testing with new

surveillance requirements to conduct

containment leak testing according to a

performance-based containment leak

test program.

Date of issuance: May 8, 1996

Effective date: May 8, 1996, to be

implemented within 30 days of

issuance.

Amendment No.: 144

Facility Operating License No. NPF21: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: February 14, 1996 (61 FR

5820) The March 19, 1996,

supplemental letter provided additional

clarifying information and did not

change the initial no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated May 8, 1996.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Richland Public Library, 955

Northgate Street, Richland, Washington

99352

Dated at Rockville, Maryland, this 15th day

of May 1996.

For the Nuclear Regulatory Commission

Steven A. Varga,

Director, Division of Reactor Projects - I/II,

Office of Nuclear Reactor Regulation

[Doc. 96-12691 Filed 5-21-96; 8:45 am]

BILLING CODE 7590-01-F

OFFICE OF THE UNITED STATES

TRADE REPRESENTATIVE

Notice of Industry Policy and Sector/

Functional Advisory Committee

Meetings

AGENCY: Office of the United States

Trade Representative.

ACTION: Notice of Industry Policy and

Sector/Functional Advisory Committee

meetings.

SUMMARY: The meetings will include a

review and discussion of current issues

which influence U.S. trade policy.

Pursuant to section 2155(f)(2) of title 19

of the United States Code, the U.S.

Trade Representative has determined

that these meetings will be concerned

with matters the disclosure of which

would seriously compromise the

Government’s negotiating objectives or

bargaining positions. Accordingly, these

meetings will be closed to the public.

DATES: The period of March 1, 1996 to

March 1, 1998.

ADDRESSES: All meetings will be held at

the U.S. Department of Commerce, 14th

Street and Independence Avenue,

Washington, D.C. 20230, unless an

alternate site is necessary.

FOR FURTHER INFORMATION CONTACT: Ms.

Phyllis Shearer Jones, Assistant U.S.

Trade Representative for

Intergovernmental Affairs and Public

Liaison, Office of the United States

Trade Representative at (202) 395–6120

or Wendy Smith, Director of the Trade

Advisory Center, Department of

Commerce at (202) 482–3268.

Charlene Barshefsky,

Acting United States Trade Representative.

[FR Doc. 96–12858 Filed 5–21–96; 8:45 am]

BILLING CODE 3190–01–M

PRESIDENTIAL ADVISORY

COMMITTEE ON GULF WAR

VETERANS’ ILLNESSES

Notice of Open Meeting

AGENCY: Presidential Advisory

Committee on Gulf War Veterans’

Illnesses.

SUMMARY: This notice is hereby given to

announce an open meeting of a panel of

the Presidential Advisory Committee on

Gulf War Veterans’ Illnesses. The panel

will discuss scientific and clinical

issues related to reproductive health

and Gulf War veterans and will receive

comment from members of the public.

Dr. Joyce C. Lashof, Advisory

Committee chair, will chair this panel

meeting.

DATES: June 17, 1996, 9:30 a.m.–4:15

p.m.; June 18, 1996, 8:30 a.m.–12:30

p.m.

PLACE: Renaissance Madison Hotel, 515

Madison Street, Seattle, WA 98104.

SUPPLEMENTARY INFORMATION: The

President established the Presidential

Advisory Committee on Gulf War

Veterans’ Illnesses by Executive Order

12961, May 26, 1995. The purpose of

this Advisory Committee is to review

and provide recommendations on the

full range of government activities

associated with Gulf War veterans’

illnesses. The Advisory Committee

reports to the President through the

Secretary of Defense, the Secretary of

Health and Human Services, and the

Secretary of Veterans Affairs. Advisory

Committee members have expertise

relevant to the functions of the

Committee and are appointed by the

President from non-Federal sectors.

Tentative Agenda

Monday, June 17, 1996.

9:30 a.m. Call to order and opening

remarks

9:40 a.m. Public comment

10:40 a.m. Break

11:00 a.m. Public comment (cont.)

12:00 .m. Lunch

1:15 p.m. Biological plausibility:

teratology, ovarian toxicity, and

spermatotoxicity

2:00 p.m. Reproductive toxicology,

hazard assessment, and the Gulf

War

2:45 p.m. Break

3:00 p.m. Epidemiology of infertility,

subfertility, fetal loss, and birth

defects in the U.S.

3:35 p.m. Evaluating rates of

congenital anomalies in children of

Gulf War veterans

4:15 p.m. Recess

Tuesday, June 18, 1996

8:30 a.m. Call to order