13738 Federal Register / Vol. 60, No. 49 / Tuesday, March 14, 1995 / Notices

of the human environment (60 FR

11124).

For further details with respect to the

action see (1) the application for

amendment dated December 22, 1993,

(2) Amendment No. 184 to License No.

DPR–61, (3) the Commission’s related

Safety Evaluation, and (4) the

Commission’s Environmental

Assessment. All of these items are

available for public inspection at the

Commission’s Public Document Room,

the Gelman Building, 2120 L Street,

NW., Washington, DC, and at the local

public document room located at the

Russell Library, 123 Broad Street,

Middletown, CT 06457.

Dated at Rockville, Maryland, this 6th day

of March 1995.

For the Nuclear Regulatory Commission.

Alan B. Wang,

Project Manager, Project Directorate I–4,

Division of Reactor Projects—I/II, Office of

Nuclear Reactor Regulation.

[FR Doc. 95–6209 Filed 3–13–95; 8:45 am]

BILLING CODE 7590–01–M

[Docket No. 50–413]

Duke Power Company, et al. (Catawba

Nuclear Station, Unit No. 1); Exemption

I

The Duke Power Company, et al. (DPC

or the licensee) is the holder of Facility

Operating License No. NPF–35, which

authorizes operation of the Catawba

Nuclear Station, Unit No. 1 (the facility),

at a steady-state reactor power level not

in excess of 3411 megawatts thermal.

The facility is a pressurized water

reactor located at the licensee’s site in

York County, South Carolina. The

license provides, among other things,

that the Catawba Nuclear Station 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) at

approximately equal intervals during

each 10-year service period of the

primary containment. The third test of

each set shall be conducted when the

plant is shut down for a 10-year

inservice inspection of the primary

containment.

III

By letters dated October 18, 1994, and

February 7, 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 1995 refueling outage, which began

on February 11, 1995, to the end-ofcycle 9 (EOC–9) refueling outage,

currently scheduled for June 1996) and

would permit the third Type A test of

the second 10-year inservice inspection

period to not correspond with the end

of the current inservice inspection

interval.

The licensee’s request concluded that

the proposed change, a one-time

extension of the interval between the

second and third ILRTs at Catawba Unit

1, is justified for the following reasons:

The previous testing history at

Catawba Unit 1 provides substantial

justification for the proposed test

interval extension. In each of the two

previous periodic ILRTs at Catawba

Unit 1, the as-found leakage was less

than or equal to 22.5% of the allowable

leakage, thereby demonstrating that

Catawba Unit 1 is a low-leakage

containment. There are no mechanisms

which would adversely affect the

structural integrity of the containment,

or that would be a factor in extending

the test interval by 20 months. However,

as a preventative maintenance measure,

a containment civil inspection,

currently required by Appendix J prior

to a Type A test, will be performed

during EOC–8 to verify that no

structural degradation exists. Any

additional risk created by the longer

interval between ILRTs is considered to

be negligible, primarily because Type B

and C testing will continue unchanged.

Additionally, the licensee stated that

its exemption request meets the

requirements of 10 CFR 50.12,

paragraphs (a)(1) and (a)(2)(ii), for the

following reasons:

In order to justify the granting of an

exemption to the requirements of 10 CFR Part

50, paragraph 50.12(a)(1) requires that the

licensee show that the proposed exemption

will not pose an undue risk to the public.

That this proposed change will not pose an

undue risk is demonstrated by the analysis

presented in draft NUREG–1493, which

concludes that an increase in the test interval

to once every 20 years would ‘‘lead to an

imperceptible increase in risk.’’ The analyses

in draft NUREG–1493 are considered to be

specifically applicable to Catawba because:

(1) The requested exemption would result in

a one-time increase in the test interval to 5

years, not 20; (2) the population density

around Catawba is less than that used in the

study (329 people per square mile, vs. 340

used in the study; (3) no ILRT at Catawba has

jailed; (4) the core inventory used in the

study was represented by a 3412 Mwt PWR

[pressurized water reactor]. Catawba is a

3411 Mwt PWR. Other factors which lead to

the conclusion that the proposed change will

not pose an undue risk include the fact that

local leak rate testing, which identifies 97%

of leakage in excess of prescribed limits, will

remain in place at its current test frequency;

the detailed, proceduralize containment civil

inspection which is normally performed in

conjunction with an ILRT will be performed

in place of the scheduled ILRT, to identify

potential structural deteriorations; and the

historical leak-tightness of the containment

structure, as evidenced by two successive

ILRTs in which the as-found leakage did not

exceed 22.5% of the allowable leakage rate.

A comparison was made between the risk

analysis presented in draft NUREG–1493 and

a probabilistic risk assessment performed for

Catawba Nuclear Station. While the

quantitative results of the NUREG are not

directly applicable to plants not used in the

study, similar conclusions can be made

concerning Catawba. NUREG–1493 indicates

that reactor accident risks are dominated by

accident sequences that result in failure or

bypass of the containment. This conclusion

is also valid for Catawba. Considering only

the Catawba accident sequences that do not

result in containment failure, containment

leakage contributes approximately 0.08 to

0.09 percent to off-site risk (whole-body

person-rem, thyroid nodules, and latent

fatalities). NUREG–1493 indicated that

containment leakage contributed from 0.02 to

0.10 percent to latent cancer risk. The

comparison between the analysis of NUREG–

1493 and the Catawba PRA concludes that

increases in containment leakage at Catawba

are expected to produce increases in accident

risk similar to the results in NUREG–1493.

Special circumstances, as defined in 10

CFR [50].12(a)(2)(ii), are present in that the

requirement to perform the third ILRT during

the ISI outage is not necessary to achieve the

underlying purpose of the rule. The purposes

of the rule, as stated in Section I of Appendix

J, are to ensure that: (a) Leakage through the

primary reactor containment and systems

and components penetrating containment

shall not exceed allowable values, and (b)

periodic surveillance of reactor containment

penetrations and isolation valves is

performed so that proper maintenance and

repairs are made. One of the significant

factors in assuring that the proposed

exemption will not pose an undue risk to the

public, as noted above, is the local leak rate

testing (LLRT) which is performed. That the

LLRT program at Catawba provides an

effective mechanism for maintaining

containment integrity is perhaps best

demonstrated by the fact that the most recent

ILRT at Catawba Unit 1 was performed at the

front end of the refueling outage; before any

repairs or adjustments were made to valves

or penetrations. Nevertheless, the as-found

leakage did not exceed 22.5% of the

allowable leakage rate. The fact that no

leakage paths were identified by an ILRT,

and that the ILRT met the acceptance criteria

with significant margin confirms the results

of the Type B and C testing.

The frequency and scope of the Type B and

C LLRT program are not being changed by

this exemption request. The LLRT program
Federal Register / Vol. 60, No. 49 / Tuesday, March 14, 1995 / Notices 13739

will continue to effectively detect

containment leakage resulting from the

degradation of active containment isolation

components, as well as containment

penetrations. Administrative limits have

been established for each Type B or C

component at a fraction of the allowable leak

rate, such that any leakage detected in excess

of the administrative limit will indicate a

potential valve or penetration degradation. In

instances in which a component’s leakage

exceeds its administrative limit,

proceduralized controls in the test program

require that a work order be written to repair

the component.

IV

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

CFR Part 50 states that a set of three

Type A leakage rate tests shall be

performed at approximately equal

intervals during each 10-year service

period.

The licensee proposes an exemption

to this section which would provide a

one-time interval extension for the Type

A test by approximately 16 months. The

Commission has determined that,

pursuant to 10 CFR 50.12(a)(1), this

exemption is authorized by law, will not

present an undue risk to the public

health and safety, and is consistent with

the common defense and security. The

Commission further determined, for the

reasons discussed below, that special

circumstances, as provided in 10 CFR

50.12(a)(2)(ii), are present justifying the

exemption; namely, that application of

the regulation in the particular

circumstances is not necessary to

achieve the underlying purpose of the

rule. 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. The NRC staff

has noted that the licensee has a good

record of ensuring a leak tight

containment. All Type A tests have

passed with significant margin and the

licensee has noted that the results of the

Type A testing have been confirmatory

of the Type B and C tests which will

continue to be performed. The licenses

has stated that it will continue to

perform the general containment civil

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 NRC staff has also made use of 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 with previous NRC

staff studies which show that Type B

and C testing can detect a very large

percentage of containment leaks. the

Catawba Unit 1 experience has also

been consistent with this.

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).

Based on generic and plant-specific

data, the NRC staff finds the licensee’s

proposed one-time exemption to permit

a schedular extension of one cycle for

the performance of the Appendix Type

A test to be acceptable.

Pursuant to 10 CFR 51.32, the

Commission has determined that

granting this exemption will not have a

significant impact on the human

environment (60 CFR 11125).

This exemption is effective upon

issuance and shall expire at the

completion of the 1996 refueling outage.

Dated at Rockville, Maryland, this 7th day

of March 1995.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Acting Director, Division of Reactor Projects

— I/II, Office of Nuclear Reactor Regulation.

[FR Doc. 95–6205 Filed 3–13–95; 8:45 am]

BILLING CODE 7590–01–M

Forrest L. Roudebush, Kansas City,

Missouri; Order Prohibiting

Involvement in NRC-Licensed

Activities and Requiring Certain

Notification to NRC

I

Mr. Forrest L. Roudebush has been,

from its inception, the owner and

president of Piping Specialists

Incorporated (PSI or Licensee), also

known as PSI Inspection, which was the

holder of Byproduct Material License

No. 24–24826–01 issued by the U.S.

Nuclear Regulatory Commission (NRC

or Commission) pursuant to 10 CFR

Parts 30 and 34 on March 6, 1987. The

license authorized the use of byproduct

material (iridium-192 and cobalt-60) for

industrial radiography in devices

approved by the NRC or an Agreement

State. The facility where licensed

materials were authorized for storage

was located at 1010 East 10th Street,

Kansas City, Missouri. The use of

licensed materials was authorized at

temporary job sites anywhere in the

United States that the NRC maintains

jurisdiction for regulating the use of

licensed materials. On October 17, 1991,

the NRC staff issued an Order

Suspending License (Effective

Immediately) to PSI. On April 22, 1992,

the NRC staff issued to PSI an Order

Modifying Order Suspending License

(Effective Immediately) and Order

Revoking License. The revocation of the

license was upheld by a decision of the

NRC Atomic Safety and Licensing Board

(ASLB), Piping Specialists, Inc. and

Forrest L. Roudebush, LBP 92–25, 36

NRC 156 (1992), which the Commission

declined to review, CLI–92–16, 36 NRC

351 (1992).

II

NRC Region III initiated an inspection

of the Licensee on September 4, 1991,

and on September 24, 1991, the NRC

Office of Investigations (OI) commenced

an investigation based on information

received on August 29, 1991, that the

PSI radiation safety program was not

being conducted in compliance with

NRC rules, regulations, and license