Federal Register / Vol. 62, No. 244 / Friday, December 19, 1997 / Notices 66699

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:

Rulemakings and Adjudications Staff, or

may be delivered to the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, by the above date. A

copy of the petition should also be sent

to the Office of the General Counsel,

U.S. Nuclear Regulatory Commission,

Washington, DC 20555–0001, and to

Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza,

Chicago, Illinois 60603, attorney for the

licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

presiding Atomic Safety and Licensing

Board that the petition and/or request

should be granted based upon a

balancing of the factors specified in 10

CFR 2.714(a)(1)(i)–(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment dated June 17, 1997, 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 rooms: for Byron,

located at the Byron Public Library

District, 109 N. Franklin, P.O. Box 434,

Byron, Illinois 61010; for Braidwood,

the Wilmington Public Library, 201 S.

Kankakee Street, Wilmington, Illinois

60481.

Dated at Rockville, Maryland, this 12th day

of December, 1997.

For the Nuclear Regulatory Commission.

George Dick, Jr.,

Project Manager Project Directorate III–2

Division of Reactor Projects—III/IV Office of

Nuclear Reactor Regulation.

[FR Doc. 97–33230 Filed 12–18–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. STN 50–454, STN 50–455, STN

50–456, and STN 50–457]

Commonwealth Edison Company;

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing Byron

Station, Units 1 and 2 and Braidwood

Station, Units 1 and 2

The U.S. Nuclear Regulatory

Commission (the Commission) is

considering issuance of an amendment

to Facility Operating License Nos. NPF–

37, NPF–66, NPF–72, and NPF–77

issued to Commonwealth Edison

Company (the licensee) for operation of

the Byron Station, Units 1 and 2, located

in Ogle County, Illinois and Braidwood

Station, Units 1 and 2, located in Will

County, Illinois.

The proposed amendment would

revise technical specification (TS) 1.0,

‘‘Definitions’’, TS 3/4.6.1, ‘‘Primary

Containment’’ and associated Bases; and

TS 5.4.2, ‘‘Reactor Coolant System

Volume’’ for Byron and Braidwood to

support the steam generator

replacement for Unit 1 at each site. The

replacement steam generators increase

the reactor coolant system volume

which results in a higher calculated

peak containment pressure (Pa) value.

The staff’s proposed no significant

hazards consideration determination for

the requested change was published on

April 23, 1997 (62 FR 19826).

Before issuance of the proposed

license amendment, the Commission

will have made findings required by the

Atomic Energy Act of 1954, as amended

(the Act) and the Commission’s

regulations.

The Commission has made a

proposed determination that the

amendment request involves 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. 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.

Each of the RSGs has a larger RCS primary

side volume than the original steam

generators (OSGs). As a result of the RCS

volume increase, the mass and energy release

during the blowdown phase of the large

break loss of coolant accident (LBLOCA) is

increased. Additionally, the heat transfer rate

of the RSGs is greater than the OSGs, and the

RSGs will operate at a slightly higher

pressure than that for the OSGs.

Consequently, the steam enthalpy exiting the

break during the reflood period, for the RSGs,

will be greater than for the OSGs. This results

in an increase in the containment building

peak pressure, Pa.

The proposed revisions to the Technical

Specifications involve the corrected value of

the current Unit 1 and Unit 2 RCS volume

and the incremental change in RCS volume

for the RSGs. The proposed revisions also

involve the defined value of Unit 1 Pa

following installation of the RSGs. Several

editorial changes are also being made to

improve clarity and consistency of the TS.

RCS volume is not an initiator for any

event and an increase in volume does not

affect any operating margin or requirements.

Therefore, increasing the primary volume

does not increase the probability of any event

previously analyzed.

The current value of Pa for Unit 2 is

unchanged due to conservatism in the

original analysis. The revised value of Pa for

Unit 1 continues to be less than the design

basis pressure for the containment structure.

The change represents only a revision to the

containment test pressure for containment

leakage testing. Such testing is only

performed with the affected unit in the

shutdown condition. Therefore, the proposed

change in Pa for Unit 1 does not involve a

significant increase in the probability of an

accident previously evaluated.

All accidents in the Updated Final Safety

Analysis Report (UFSAR) were evaluated to

determine the effect of an increase in primary

volume on accident consequences. The

events identified that may be impacted by an

increase in primary volume are the Waste

Gas System Leak or Failure and LBLOCA. For

the Waste Gas System Leak or Failure, the

activity of the decay tank is controlled to

Technical Specification limits which are

unaffected by RCS volume. Therefore, an

increase in RCS volume would not increase

the offsite dose.

The offsite dose calculation for the

LBLOCA is unaffected by the proposed
66700 Federal Register / Vol. 62, No. 244 / Friday, December 19, 1997 / Notices

change. The license basis offsite dose

calculation is in accordance with NRC Reg

Guide 1.4 ‘‘Assumptions Used for Evaluating

The Potential Radiological Consequences of a

Loss of Coolant Accident for Pressurized

Water Reactors.’’ This Regulatory Guide

states, in part, ‘‘* * * a number of

appropriately conservative assumptions,

based on engineering judgment and on

applicable experimental results from safety

research programs conducted by the AEC.’’

These conservatisms include (but are not

limited to) the following assumptions:

Twenty five percent of the equilibrium full

power radioactive iodine inventory is

immediately available for leakage from the

primary containment. 100% of the

equilibrium full power radioactive noble gas

inventory is immediately available for

leakage from the primary containment. The

primary containment should be assumed to

leak at the (maximum) leak rate specified in

the technical specifications for the first 24

hours and at 50% of this value for the

remaining 29 days of the accident duration.

The design basis leakage corresponding to

a peak containment pressure of 50 psig

utilized in the design basis accident analysis

is 0.10% per day of the containment free air

mass. Therefore, the offsite dose calculation

was performed with a leakage of .1% per day

for day one and .05% per day for days 2

through 30. Isotopic inventories are

unaffected by the increase in reactor coolant

volume. Thus, the offsite dose is unaffected

by the increase in the peak containment

pressure. Therefore, this proposed change to

Pa does not involve a significant increase in

the consequences of an accident previously

evaluated.

The editorial changes proposed are for

clarity and consistency within the Technical

Specifications and do not affect either the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change in RCS volume is a

change in a plant parameter within the

‘‘Design Features’’ section of the Technical

Specifications. Increasing the RCS volume

does not create any new or different failure

modes. The existing RCS design

requirements continue to be met.

The revised value of Pa for Unit 1 following

replacement of steam generators continues to

be less than the design basis pressure for the

containment building structure. The change

represents only a revision to the test pressure

for containment leakage testing. Such testing

is only performed with the affected unit in

the shutdown condition. Therefore, no new

or different failure modes are being

introduced by modification of the testing

parameters.

The editorial changes proposed are for

clarity and consistency within the Technical

Specifications and do not result in any

physical changes to the facility or how it is

operated. No new or different failure modes

are being introduced by these changes.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

Changing the RCS volume in the Technical

Specifications does not reduce the margin of

safety. RCS volume is a design feature. An

evaluation of all UFSAR accidents was

performed to determine the effect of an

increase in RCS volume. This evaluation is

summarized as follows:

An evaluation of the Chemical and Volume

Control System Malfunction was performed

to determine the effect of the increased RCS

volume. The larger RCS volume reduces the

reactivity insertion for a given dilution flow

rate. Therefore, the UFSAR analyses remain

bounding for Byron and Braidwood and there

is no reduction in the margin of safety.

An evaluation of the Inadvertent Actuation

of the Emergency Core Cooling System

During Power Operation Event was

performed to determine the effect of the

increased RCS volume due to the RSGs. For

this event, the injection of borated water

causes a negative reactivity insertion, which

increases DNBR. For a given Refueling Water

Storage Tank (RWST) boron concentration,

the larger RCS volume will cause a reduction

in the negative reactivity insertion rate as

compared to the current UFSAR analysis.

However, negative reactivity would still be

inserted and no fuel pins would experience

DNB. Additionally, the increased RCS

volume was evaluated to determine the effect

on pressurizer level following the inadvertent

actuation of ECCS and was found to be

acceptable. Therefore, there is no reduction

in the margin of safety.

An evaluation of the Small Break LOCA

was performed to determine the effect of

increased RCS volume. The additional RCS

volume will cause a delay in the loop seal

clearing which in turn delays the core

uncovery as compared with the UFSAR

analysis. A delay in core uncovery reduces

the amount of core heatup which results in

a lower peak clad temperature (PCT) because

the core decay heat would be less than in the

UFSAR analysis. The benefit is considered

small, but there is still a benefit. Therefore,

the increased RCS volume does not result in

a reduction in the margin of safety.

An evaluation of the Large Break LOCA

was performed to determine the effect of

increased RCS volume for the RSGs. For a LB

LOCA, the increased RCS volume causes the

blowdown phase of the event to be longer.

Increased blowdown phase, alone, could

potentially result in a higher PCT. However,

the RSGs also have less resistance to flow

due to increased primary side steam

generator flow area, which results in a higher

blowdown flow compared to the OSGs. The

increased blowdown flow will compensate

for the longer blowdown phase associated

with the increased RCS volume. The net

effect is that the blowdown time (end of

bypass) for the RSG will be the same or

decrease compared to the OSG. Reduced

resistance to break flow for the RSG

compared to the OSG will result in a lower

PCT for the RSG compared to the OSG.

The increase in the current value of RCS

volume in Unit 2 is significantly less than the

increase associated with the replacement of

the steam generators in Unit 1. The small

increase in the RCS volume will likely result

in a slight increase in the blowdown period.

This slight increase in the blowdown period

will have no significant impact on the peak

clad temperature (PCT) calculation for Unit

2. Any small changes in the PCT due to this

small increase in the RCS volume can be

easily accommodated for Unit 2 because of

the significant margin in the PCT (over 100

degrees) available to the Appendix K 10 CFR

50.46 acceptance criteria of 2200 °F.

Therefore, there is no reduction in the margin

of safety.

An evaluation of the Gas Waste System

Leak or Failure was performed to determine

the effect of the increased RCS volume.

Because the activity of the decay tank is

controlled within Technical Specification

limits, an increase in RCS volume would not

change the results of the event. Therefore,

there is no reduction in the margin of safety.

An evaluation was performed to determine

the effect of the increased RCS volume

(associated with the RSGs) on the peak

containment pressure following a LBLOCA.

The increased RCS volume caused the peak

containment pressure to increase to 47.8 psig.

This is still below the containment design

pressure of 50.0 psig. Therefore, there is no

reduction in the margin of safety. The

increase in RCS volume for the existing units

(without RSGs) remains within the

conservative volume used in the calculation

of the current peak containment pressure

value of 44.4 psig. Therefore, there is no

reduction in the margin of safety.

This proposed change involves testing

requirements designed to demonstrate

acceptable leakage rates are maintained. If

acceptable leakage rates are maintained as

outlined in the Technical Specifications,

there will be no reduction in the margin of

safety. In the event of degradation of a

containment seal that results in unacceptable

leakage, plant shutdown will occur as

required by Technical Specifications and

administrative requirements in accordance

with approved plant procedures. Therefore,

this proposed change does not involve a

significant reduction in a margin of safety.

The editorial changes proposed are for clarity

and consistency within the Technical

Specifications and do not result in any

physical changes to the facility or how it is

operated. Therefore, the changes have no

effect on the margin of safety.

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
Federal Register / Vol. 62, No. 244 / Friday, December 19, 1997 / Notices 66701

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received. 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 hearing and

petitions for leave to intervene is

discussed below.

By January 20, 1998, the licensee may

file a request for a hearing with respect

to issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room located for Byron, the

Byron Public Library District, 109 N.

Franklin, P.O. Box 434, Byron, Illinois

61010; for Braidwood, the Wilmington

Public Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481. 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 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 party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

Rulemakings and Adjudications Staff, or

may be delivered to the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, by the above date. A

copy of the petition should also be sent

to the Office of the General Counsel,

U.S. Nuclear Regulatory Commission,

Washington, DC 20555–0001, and to

Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza,

Chicago, Illinois 60603, attorney for the

licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

presiding Atomic Safety and Licensing

Board that the petition and/or request

should be granted based upon a

balancing of the factors specified in 10

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

For further details with respect to this

action, see the application for

amendment dated January 30, 1997, as

revised on December 9, 1997, 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 rooms: for Byron,

located at the Byron Public Library

District, 109 Franklin, P.O. Box 434,

Byron, Illinois 61010; for Braidwood,

the Wilmington Public Library, 201 S.

Kankakee Street, Wilmington, Illinois

60481.

Dated at Rockville, Maryland, this 15th day

of December, 1997.
66702 Federal Register / Vol. 62, No. 244 / Friday, December 19, 1997 / Notices

For the Nuclear Regulatory Commission.

George Dick, Jr.,

Project Manager, Project Directorate III–2,

Division of Reactor Projects—III/IV, Office of

Nuclear Reactor Regulation.

[FR Doc. 97–33231 Filed 12–18–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Policy Statement on Cooperation With

States at Commercial Nuclear Power

Plants and Other Production or

Utilization Facilities: Notice of

Approval

On February 22, 1989 (54 FR 7530) as

revised on February 25, 1992 (57 FR

6462), the NRC published a policy

statement addressing cooperation

between the NRC and States concerning

commercial nuclear power plants and

other utilization facilities. The NRC has

received a renewal of the clearance from

the Office of Management and Budget

(OMB) for the information collection

requirements under the provisions of

the Paperwork Reduction Act of 1995,

Public Law 104–13 (44 U.S.C. Chapter

35). The policy statement is approved

under OMB control number 3150–0163.

Dated at Rockville, Maryland this 15th day

of December, 1997.

For the Nuclear Regulatory Commission.

Brenda Jo. Shelton,

NRC Clearance Officer, Office of the Chief

Information Officer.

[FR Doc. 97–33220 Filed 12–18–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

State of New Mexico Relinquishment of

Sealed Source and Device Evaluation

and Approval Authority and

Reassumption by the Commission

AGENCY: Nuclear Regulatory

Commission.

ACTION: Notice of reassumption of sealed

source and device evaluation and

approval authority from the State of

New Mexico.

SUMMARY: Notice is hereby given that

effective January 1, 1998, the Nuclear

Regulatory Commission will reassume

regulatory authority for sealed source

and device evaluations and approvals in

the Agreement State of New Mexico in

response to a request from the Governor

of the State of New Mexico to relinquish

this authority.

EFFECTIVE DATE: January 1, 1998.

FOR FURTHER INFORMATION CONTACT: Ms.

Cardelia H. Maupin, Senior Project

Manager, Office of State Programs, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, telephone (301)

415–2312, Internet: CHM@NRC.GOV.

SUPPLEMENTARY INFORMATION: Currently,

the State of New Mexico has an

Agreement with the Nuclear Regulatory

Commission (NRC) which grants the

State authority to regulate specific

categories of radioactive materials

formerly regulated by the NRC. This

Agreement was entered into on May 1,

1974, pursuant to Section 274b of the

Atomic Energy Act of 1954, as amended.

Recently, the NRC received a letter

from New Mexico Governor Gary E.

Johnson (September 8, 1997) requesting

relinquishment of the State’s authority

to evaluate and approve sealed source

and devices, and assumption of this

authority by NRC. The requested action

would involve reassertion of regulatory

authority by NRC over activities

currently regulated by New Mexico

pursuant to its Agreement with NRC.

The Governor of New Mexico noted

there are two manufacturers in the State

and there has been no sealed source and

device evaluations conducted since

1988. Governor Johnson indicated that it

would not be cost effective to fund and

maintain staff to conduct sealed source

and device evaluations.

The Commission has agreed to the

request and has notified New Mexico

that effective January 1, 1998, the NRC

will reassume authority to evaluate and

approve sealed source and device

applications within the State of New

Mexico. The State of New Mexico will

retain authority to regulate the

manufacture and use of sealed sources

and devices within the State in

accordance with its Section 274b

Agreement with the NRC.

Dated at Rockville, Maryland this 12th day

of December, 1997.

For the Nuclear Regulatory Commission.

John C. Hoyle,

Secretary of the Commission.

[FR Doc. 97–33218 Filed 12–18–97; 8:45 am]

BILLING CODE 7590–01–P

RAILROAD RETIREMENT BOARD

Privacy Act of 1974, Proposed

Changes to System of Records

AGENCY: Railroad Retirement Board.

ACTION: Notice of a proposed routine

use.

SUMMARY: The purpose of this document

is to give notice of a proposed routine

use to one of the RRB’s Privacy Act

systems of records.

DATES: The new routine use will be

effective 30 calendar days from the date

of this publication unless comments are

received before this date which would

result in a contrary determination.

ADDRESSES: Send comments to Beatrice

Ezerski, Secretary to the Board, Railroad

Retirement Board, 844 North Rush

Street, Chicago, Illinois 60611–2092.

FOR FURTHER INFORMATION CONTACT:

LeRoy Blommaert, Privacy Act Officer,

Railroad Retirement Board, 844 North

Rush Street, Chicago, Illinois 60611–

2092, (312) 751–4548.

SUPPLEMENTARY INFORMATION:

I. Discussion of Proposed Routine Use

Pursuant to Public Law 104–193, the

Personal Responsibility and Work

Opportunity Reconciliation Act of 1996,

the Railroad Retirement Board (RRB)

will disclose data from its system of

records RRB–19, Payroll Record System,

to the Office of Child Support

Enforcement, Administration for

Children and Families, Department of

Health and Human Services (HHS) for

use in the National Database of New

Hires, part of the Federal Parent Locator

System (FPLS) and Federal Tax Offset

System, DHHS/OCSE No. 09–90–0074.

A description of the Federal Parent

Locator Service may be found at 62 FR

51663 (October 2, 1997).

FPLS is a computerized network

through which States may request

location information from Federal and

State agencies to find non-custodial

parents and/or their employers for

purposes of establishing paternity and

securing support. On October 1, 1997,

the FPLS was expanded to include the

National Directory of New Hires, a

database containing information on

employees recently hired, quarterly

wage data on private and public sector

employees, and information on

unemployment compensation benefits.

On October 1, 1988, the FPLS will be

expanded further to include a Federal

Case Registry. The Federal Case Registry

will contain abstracts on all participants

involved in child support enforcement

cases. When the Federal Case Registry is

instituted, its files will be matched on

an ongoing basis against the files in the

National Directory of New Hires to

determine if an employee is a

participant in a child support case

anywhere in the country. If the FPLS

identifies a person as being a participant

in a State child support case, that State

will be notified. State requests to the

FPLS for location information will also

continue to be processed after October

1, 1998.

When individual are hired by the

RRB, we may disclose to the FPLS their