17090 Federal Register / Vol. 60, No. 64 / Tuesday, April 4, 1995 / Notices

FOR FURTHER INFORMATION CONTACT:

Daniel Solomon, International Child

Labor Study, Bureau of International

Labor Affairs, Room S–1308, U.S.

Department of Labor, Washington, DC

20210, telephone: (202) 501–6068; fax

(202) 219–4923. Persons with

disabilities who need special

accommodations should contact Mr.

Solomon by April 24, 1995.

Additional Information

The Senate Appropriations

Committee report states:

Child labor is a silent and tragic

emergency of our time. Few human

rights abuses are so unanimously

condemned, while being so universally

practiced, as child labor. The number of

children working, and the scale of their

suffering, increases year by year.

UNICEF and the International Labor

Organization estimate that hundreds of

millions of children are working today,

many in servitude and under hazardous

conditions.

Therefore, the Committee [directs the

Secretary] to continue and expand

efforts by the Department to identify

foreign industries and their host

countries that utilize child labor in the

production of goods from industry,

plantations, and mining exported to the

United States.

The Secretary is directed to utilize all

available information, including

information made available by UNICEF,

the International Labor Organization

and human rights organizations and

report his findings to the Committee no

later than July 30, 1995.

All written or oral comments

submitted pursuant to the public

hearing will be made part of the record

of review referred to above and will be

available for public inspection.

Signed at Washington, DC, this 29th day of

March 1995.

Joaquin F. Otero,

Deputy Under Secretary.

[FR Doc. 95–8225 Filed 4–3–95; 8:45 am]

BILLING CODE 4510–28–M

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–339]

Virginia Electric and Power Co. (North

Anna Power Station Unit No. 2);

Exemptions

I

Virginia Electric and Power Company

(the licensee) is the holder of Facility

Operating License No. NPF–7, which

authorizes operation of North Anna

Power Station, Unit 2 (the facility or

NA–2), at a steady-state reactor power

level not in excess of 2893 megawatts

thermal. The facility is a pressurized

water reactor located at the licensee’s

site in Louisa County, Virginia. The

license provides among other things,

that it is subject to all rules, regulations,

and Orders of the U.S. Nuclear

Regulatory Commission (the

Commission or NRC) now or hereafter

in effect.

II

Section III.D.1.(a) of appendix J to 10

CFR part 50 requires the performance of

three Type A containment integrated

leakage rate tests (ILRTs) of the primary

containment, at approximately equal

intervals during each 10-year service

period. The third test of each set shall

be conducted when the plant is shut

down for the 10-year inservice

inspection program.

Section IV.A of appendix J to 10 CFR

part 50 requires that any modification,

replacement of a component which is

part of the primary reactor containment

boundary, or resealing a seal-welded

door, performed after the preoperational

leakage rate test shall be followed by

either a Type A, Type B, or Type C test,

as applicable for the area affected by the

modification.

III

By letter dated March 2, 1995, the

licensee requested temporary relief from

the requirement to perform a set of three

Type A tests at approximately equal

intervals during each 10-year service

period of the primary containment. The

requested exemption would permit a

one-time interval extension of the third

Type A test by approximately 16

months (from the March 1995 steam

generator replacement outage, to the

October 1996 refueling outage).

The licensee’s March 2, 1995, letter

also requested temporary relief from the

requirements to perform a type A test

following a major modification or

replacement of a component which is

part of the primary reactor coolant

boundary. Specifically, the postmodification exemption is requested

from performing a Type A test due to

the activities associated with the

upcoming NA–2 steam generator

replacement. The basis for the postmodification exemption request is that,

in this case, the ASME Section XI

inspection and testing requirements

more than fulfill the intent of the

requirements of Section IV.A of

Appendix J.

The licensee’s request cites the

special circumstances of 10 CFR 50.12,

paragraph (a)(2)(ii), as the basis for the

exemption to Section III.D.1.a of

appendix J to 10 CFR part 50. The

licensee points out that the existing

Type B and C testing programs are not

being modified by this request and will

continue to effectively detect

containment leakage caused by the

degradation of active containment

isolation components as well as

containment penetrations. It has been

the experience at NA–2 during the Type

A tests conducted during the first 10-

year inservice inspection interval (1984,

1989, and 1990), that considerable

margin exists between the Type A tests

and the Technical Specifications (TS)

allowable leakage rate limit.

During operation, the NA–2

containment is maintained at a

subatmospheric pressure

(approximately 10.0 psia) which

provides a good indication of the

containment integrity. TS require the

containment to be subatmospheric when

in Modes 4, 3, 2, and 1. Containment air

partial pressure is monitored in the

control room to ensure TS compliance.

If the containment air partial pressure

increases above the established TS limit,

the unit is required to shut down.

The licensee’s request also cites the

special circumstances of 10 CFR 50.12,

paragraph (a)(2)(ii), as the basis for the

exemption to Section IV.A of appendix

J to 10 CFR part 50.

The NA–2 plant design incorporates a

‘‘closed system’’ for transferring steam

from the steam generators inside of the

primary containment to the main

turbine-generators in the turbine

building. The inside containment

portion of this closed system consists of

the main steam lines, the feedwater

lines, and the secondary side of the

steam generators. This closed system

inside of containment forms a part of

the primary reactor containment

boundary.

The planned replacement of the NA–

2 steam generators includes the

following activities:

—Cutting and removing the mainsteam

and feedwater lines from the steam

generators.

—Cutting and removing the upper

assemblies of the steam generators

(steam domes).

—Cutting the reactor coolant piping and

removing the steam generator lower

assemblies (tube bundles).

—Installing the new steam generator

lower assemblies and re-welding the

reactor coolant piping.

—Re-installing the steam generator

upper assemblies on the new lower

assemblies.

—Re-installing and re-welding the main

steam and feedwater lines.

The planned replacement of the NA–

2 steam generators affects only this
Federal Register / Vol. 60, No. 64 / Tuesday, April 4, 1995 / Notices 17091

closed piping system inside

containment. The steam generator

replacement activities do not affect the

containment structure or the actual

containment liner.

Section IV.A to Appendix J, Special

Testing Requirements for Containment

Modifications, requires that any major

modification or replacement of a

component which is part of the primary

reactor containment boundary shall be

followed by either a Type A, Type B, or

Type C test, as applicable for the area

affected by the modification. The Type

C testing requirements of Appendix J

apply to leakage testing of containment

isolation valves. The planned

replacement does not affect any

containment isolation valves and,

therefore, the Type C testing

requirements are not applicable. The

Type B testing requirements of

appendix J apply to leakage testing of

gasketed or sealed containment

penetrations (e.g., electrical

penetrations), air lock door seals, and

other doors with resilient seals or

gaskets. Although the secondary side of

the steam generators have access

manways with gaskets, the Type B

testing requirements do not address the

other areas of the containment boundary

affected by the planned replacement,

i.e., weld seams in the steam generator

and in the main steam and feedwater

piping. Hence, because the affected

areas cannot be tested by Type B or

Type C testing, Section IV.A of

Appendix J would require that a Type

A test be performed prior to startup

following the planned steam generator

replacement.

However, the affected area of the

primary containment boundary is also

part of the pressure boundary of an

ASME Class 2 component/piping

system and, as such, the planned

replacement of the steam generators is

subject to the repair and replacement

requirements of ASME Section XI. The

ASME Section XI surface examination,

volumetric examination, and system

pressure test requirements are more

stringent than the Type A testing

requirements of Appendix J. The

acceptance criteria for ASME Section XI

system pressure testing of welded joints

is ‘‘zero leakage.’’ In addition, the test

pressure for the system pressure test

will be in excess of 20 times that of a

Type A test (1356 psig vs. 44.1 spig).

Therefore, the ASME Section XI

inspection and testing requirements

more than fulfill the intent of the

requirements of Section IV.A of

appendix J.

IV

In the licensee’s March 2, 1995,

exemption request, the licensee stated

that special circumstance 50.12(a)(2)(ii)

is applicable to this situation, i.e., that

application of the regulation is not

necessary to achieve the underlying

purpose of the rule.

Appendix J states that the leakage test

requirements provide for periodic

verification by tests of the leak tight

integrity of the primary reactor

containment. Appendix J further states

that the purpose of the tests ‘‘is to assure

that leakage through the primary reactor

containment shall not exceed the

allowable leakage rate values as

specified in the Technical

Specifications or associated bases.’’

Thus, the underlying purpose of the

requirement to perform Type A

containment leak rate tests at intervals

during the 10-year service period is to

ensure that any potential leakage

pathways through the containment

boundary are identified within a time

span that prevents significant

degradation from continuing or

becoming unknown.

The NRC staff has reviewed the basis

and supporting information provided by

the licensee in the exemption request

from the requirements of Section

III.D.1(a) of appendix J. The NRC staff

has noted that the licensee’s record of

ensuring a leak-tight containment has

verified containment integrity and, as

noted previously, considerable margin

exists between the Type A test results

and the TS allowable leakage rate. The

Type A tests performed in 1984, 1989,

and 1990 have all successfully verified

containment integrity. All ‘‘as-found’’

Type A test results since 1984 have been

confirmatory of the Type B and C tests

which will continue to be performed.

The licensee will perform the general

containment inspection although it is

only required by appendix J (Section

V.A.) to be performed in conjunction

with Type A tests. The NRC staff

considers that these inspections, though

limited in scope, provide an important

added level of confidence in the

continued integrity of the containment

boundary.

The NA–2 containment is of the

subatmospheric design. During

operation, the containment is

maintained at a subatmospheric

pressure (approximately 10 psia) which

provides for constant monitoring of the

containment integrity and further

obviates the need for Type A testing at

this time. If the containment air partial

pressure exceeds the established TS

limit, the unit must be shut down.

The NRC staff has also made use of

the information in a draft staff report,

NUREG–1493, which provides the

technical justification for the present

appendix J rulemaking effort which also

includes a 10-year test interval for Type

A tests. The integrated leakage rate test,

or Type A test, measures overall

containment leakage. However,

operating experience with all types of

containments used in this country

demonstrates that essentially all

containment leakage can be detected by

local leakage rate tests (Type B and C).

According to results given in NUREG–

1493, out of 180 ILRT reports covering

110 individual reactors and

approximately 770 years of operating

history, only 5 ILRT failures were found

which local leakage rate testing could

not detect. This is 3% of all failures.

This study agrees well with previous

NRC staff studies which show that Type

B and C testing can detect a very large

percentage of containment leaks.

The Nuclear Management and

Resources Council (NUMARC), now the

Nuclear Energy Institute (NEI), collected

and provided the NRC staff with

summaries of data to assist in the

appendix J rulemaking effort. NUMARC

collected results of 144 ILRTs from 33

units; 23 ILRTs exceeded 1.0La. Of

these, only nine were not due to Type

B or C leakage penalties. The NEI data

also added another perspective. The NEI

data show that in about one-third of the

cases exceeding allowable leakage, the

as-found leakage was less than 2La; in

one case the leakage was found to be

approximately 2La; in one case the asfound leakage was less than 3La; one

case approached 10La; and in one case

the leakage was found to be

approximately 21La. For about half of

the failed ILRTs the as-found leakage

was not quantified. These data show

that, for those ILRTs for which the

leakage was quantified, the leakage

values are small in comparison to the

leakage value at which the risk to the

public starts to increase over the value

of risk-corresponding to La

(approximately 200La, as discussed in

NUREG–1493). Therefore, based on

those considerations, it is unlikely that

an extension of one cycle for the

performance of the appendix J, Type A

test at NA–2 would result in significant

degradation of the overall containment

integrity. As a result, the special

circumstances of 10 CFR 50.12(a)(2)(ii)

are present in that the application of the

regulation in these particular

circumstances is not needed to achieve

the underlying purpose of the rule.

Based on generic and plant specific

data, the NRC staff finds the basis for

the licensee’s proposed exemption to
17092 Federal Register / Vol. 60, No. 64 / Tuesday, April 4, 1995 / Notices

allow a one-time exemption to permit a

schedular extension of one cycle for the

performance of the appendix Type A

test, provided that the general

containment inspection is performed, to

be acceptable.

Section IV.A of appendix J would

normally require that a Type A test be

performed prior to startup following a

containment modification such as the

planned steam generator replacement.

However, in this case, the affected area

of the primary containment boundary is

also part of the pressure boundary of a

ASME Class 2 component/piping

system and, as such, the planned

replacement of the steam generators is

subject to the repair and replacement

requirements of ASME Section XI. The

ASME Section XI surface examination,

volumetric examination, and system

pressure testing requirements are more

stringent than the Type A testing

requirements of appendix J. The

objective of the Type A test required by

Section IV.A is to assure the leak-tight

integrity of the containment area

affected by the modification. The ASME

Section XI inspection and testing

requirements more than fulfill the intent

of the requirements of Section IV.A of

appendix J. As a result, the special

circumstances of 10 CFR 50.12(a)(2)(ii)

are present in that the application of the

regulation in these particular

circumstances is not needed to achieve

the underlying purpose of the rule.

Therefore, the NRC staff finds the basis

for the licensee’s proposed exemption to

allow a one-time exemption from Type

A testing for modification of the primary

containment boundary due to the

forthcoming NA steam generator

replacement to be acceptable.

Pursuant to 10 CFR 51.32, the

Commission has determined that

granting these Exemptions will not have

a significant impact on the environment

(60 FR 15945).

The exemption from Section III.D.1.(a)

of appendix J to 10 CFR part 50 is

effective upon issuance and shall expire

at the completion of the NA–2 1996

refueling outage.

The exemption from Section IV.A of

appendix J to 10 CFR part 50 is effective

upon issuance and shall expire at the

completion of the NA–2 1995 steam

generator replacement refueling outage.

For the Nuclear Regulatory Commission.

Dated at Rockville, MD, this 29th day of

March 1995.

Steven A. Varga,

Director, Division of Reactor Projects—I/II,

Office of Nuclear Reactor Regulation.

[FR Doc. 95–8166 Filed 4–3–95; 8:45 am]

BILLING CODE 7590–01–M

NUCLEAR WASTE TECHNICAL

REVIEW BOARD

Panel on the Engineered Barrier

System: Meeting and Tour of the Idaho

National Engineering Laboratory

Pursuant to its authority under

section 5051 of Public Law 100–203, the

Nuclear Waste Policy Amendments Act

of 1987, the Nuclear Waste Technical

Review Board’s Panel on the Engineered

Barrier System will hold a meeting on

Tuesday, June 6, 1995, in Idaho Falls,

Idaho, and a tour of the Idaho National

Engineering Laboratory (INEL) site on

Wednesday, June 7, 1995. The meeting

will be held at the Shilo Inn, 780

Lindsay Boulevard, Idaho Falls, Idaho

83402; Tel (208) 523–0088; Fax (208)

522–7420. The meeting and tour are

open to the public; however, space on

the tour is limited and advance

reservations are required.

The panel meeting on Tuesday will

focus on three areas of interest to the

Board: (1) government-owned spent

nuclear fuel at INEL (its description and

plans for its eventual permanent

disposal), (2) contaminated scrap metal

and greater-than-class-C waste activities

managed by the Department of Energy’s

Idaho Operations Office that could have

an impact on permanent disposal in a

repository, and (3) dry storage of spent

nuclear fuel, including current research

and development activities at INEL for

government-owned and commercial

spent nuclear fuel. Panel members have

invited the Department of Energy and its

INEL contractors and INEL researchers

to discuss these issues. The panel will

also hear about the status of efforts to

get INEL high-level defense wastes such

as calcine and tank-stored liquids into

appropriate forms for transportation to

and disposal in a potential repository.

On Wednesday, June 7, the penal will

participate in a tour of the INEL

facilities discussed in the previous day’s

meeting. The Board makes every effort

to ensure that the general public has

access to all of its activities. To that end,

the public is invited to attend the tour

with the panel. Space is limited,

however, and will be filled on a firstcome, first-served basis. The tour will

begin at the Shilo Inn in Idaho Falls at

approximately 8 a.m. and return to the

hotel at approximately 6 p.m.

All who wish to join the tour must

provide the following information to

Frank Randall, (703) 235–4473 or FAX

(703) 235–4495.

1. Full name

2. Social security number

3. Date of birth

4. Daytime telephone number

5. Company or organization

6. Place of birth (city and state)

7. Country of citizenship (if non-U.S.)

U.S. citizens must call or fax their

data to Mr. Randall by May 19, 1995.

Non-U.S. citizens must call or fax their

data to Mr. Randall by April 28, 1995.

No one will be registered for the tour

after the applicable cutoff date.

The Nuclear Waste Technical Review

Board was created by Congress in the

Nuclear Waste Policy Amendments Act

of 1987 to evaluate the technical and

scientific validity of activities

undertaken by the DOE in its program

to manage the disposal of the nation’s

high-level radioactive waste and spent

nuclear fuel. In that same legislation,

Congress directed the DOE to

characterize a site at Yucca Mountain,

Nevada, for its suitability as a potential

location for a permanent repository for

the disposal of that waste.

Transcripts of the meeting will be

available on computer disk or on a

library-loan basis in paper format from

Victoria Reich, Board librarian,

beginning July 24, 1995. For further

information, contact Frank Randall,

External Affairs, Nuclear Waste

Technical Review Board, 1100 Wilson

Boulevard, Suite 910, Arlington,

Virginia 22209; (703) 235–4473.

Dated: March 29, 1995.

William Barnard,

Executive Director, Nuclear Waste Technical

Review Board.

[FR Doc. 95–8146 Filed 4–3–95; 8:45 am]

BILLING CODE 6820–AM–M

PRESIDENT’S COUNCIL ON

SUSTAINABLE DEVELOPMENT

The Eighth Meeting of the President’s

Council on Sustainable Development

(PCSD) in San Francisco, CA

SUMMARY: The President’s Council on

Sustainable Development, a partnership

of industry, government, and

environmental, labor and civil rights

organizations, will convene its eighth

meeting in San Francisco, California.

Council members will further discuss

the PCSD’s role in developing

recommendations to the President

toward the integration of environmental

and economic policy and, ultimately,

establishing a long-term path toward a

sustainable United States by the year

2040.

Council members will discuss at

length the draft policy

recommendations for a sustainable

future, which have been developed by

each of the PCSD’s task forces. The task

forces have generated these policy

recommendations based on information