Federal Register / Vol. 60, No. 238 / Tuesday, December 12, 1995 / Notices 63739

In situ pressure tests were conducted

on the tubes with the largest MRPC

indications and the results indicate

acceptable margins against burst under

normal operating and postulated

accident conditions. The NRC had a

review conducted by an independent

contractor of the in situ test method

used at Maine Yankee and determined

that it provides a reasonable simulation

of the hydraulic pressure loads induced

during a postulated main steamline

break.

Thus, it has been demonstrated that

the tubes with the largest indications at

Maine Yankee continued to exhibit

adequate structural integrity at the time

they were found. This finding is

attributable to the morphology of the

cracks as determined from

metallographic examinations of pulled

tube specimens from Maine Yankee.

This morphology consists of cracks that

were not coplanar but rather of short

circumferential length and staggered

around the circumference over a short

axial region with ligaments of material

between the cracks. These ligaments

add considerably to the strength of the

tube, but these ligaments are generally

not detectable by the MRPC.

The findings at Maine Yankee

nevertheless raised concern that large

undetected circumferential cracks could

possibly exist at other plants. Therefore,

the NRC issued Generic Letter (GL) 95–

03, ‘‘Circumferential Cracking of Steam

Generator Tubes,’’ on April 28, 1995,

notifying licensees of the Maine Yankee

experience and requesting that they

evaluate recent operating experience

concerning the detection and sizing of

circumferential cracks and the potential

applicability of this experience to their

plants. On the basis of the results of this

evaluation, past inspections and the

results thereof, and other relevant

factors, licensees were requested to

develop a safety assessment justifying

continued operation until the next

scheduled steam generator tube

inspections were to be performed. The

generic letter also requested that

licensees develop and submit their

plans for the next steam generator tube

inspection as they pertain to the

detection of circumferential cracks. The

utilities were required to respond to GL

95–03 within 60 days. By now, the

utilities that own the six plants listed in

the Petition have responded to GL 95–

03 and the responses have been

evaluated by the staff.

Based on the utilities’ responses to GL

95–03, with the exception of Millstone

Unit 2, the CE plants listed in the

Petition have been inspected in those

areas susceptible to circumferential

cracking with improved eddy current

inspection probes equally capable as the

Point Plus system of detecting

circumferential cracking. All tubes with

detected cracks have been removed from

service. The licensee for Millstone Unit

2 replaced the original CE steam

generators during an outage that ended

in January 1993. The new steam

generators incorporated many new

design features that are expected to

eliminate or greatly reduce the potential

for circumferential tube cracking. These

include the use of Inconel 690, a

material that has significantly greater

resistance to cracking and hydraulic

expansion of tubes, which reduces the

potential for cracking in the expansion

transitions. The limited operational

time, improvements in design, and

favorable plant operating conditions

minimize the potential for the

development of circumferential cracking

in the Millstone Unit 2 steam

generators. Millstone Unit 2 steam

generators will continue to be inspected

during refueling outages.

The NRC has studied the risk and

potential consequences of a range of

SGTR events in NUREG–0844, ‘‘NRC

Integrated Program for the Resolution of

Unresolved Safety Issues A–3, A–4, and

A–5 Regarding Steam Generator Tube

Integrity.’’ The staff estimated the risk

contribution due to the potential for

single and multiple SGTRs. The study

also examined the expected

consequences of SGTR scenarios,

including beyond-design-basis

situations, such as the potential for

release as a result of containment bypass

because of failed tubes concurrent with

a breach of secondary system integrity.

A combination of circumstances and

conditions is required to produce such

simultaneous failures: (1) Main

steamline break or other less severe loss

of secondary system integrity, (2) the

potential that a population of tubes

susceptible to rupture exists in a

particular steam generator, (3) the

potential that operators would not take

actions to avoid high differential

pressures, and (4) the probability that a

large number of tubes would actually

fail simultaneously. In the NUREG

–0844 assessment, the staff concluded

that the probability of simultaneous

multiple tube failure was small

(approximately 10¥5), and that the risk

resulting from releases during SGTRs

with loss of secondary system integrity

was small (about 10¥7 latent fatalities

per reactor year).

III. Conclusion

Based on the fact that (1) adequate

steam generator tube inspections have

been performed, (2) primary-tosecondary leakage is being monitored on

a continuing basis, and (3) the risk of

multiple SGTR events is low, I have

concluded that an immediate shutdown

and Plus Point probe inspection of

Maine Yankee, Fort Calhoun Unit 1,

Calvert Cliffs Units 1 and 2, St. Lucie

Unit 1, and Millstone Unit 2 is not

warranted.

The Petitioner’s request for action

pursuant to 10 CFR 2.206 is denied. As

provided in 10 CFR 2.206(c), a copy of

the Decision will be filed with the

Secretary of the Commission for the

Commission’s review. This Decision

will constitute the final action of the

Commission 25 days after issuance

unless the Commission, on its own

motion, institutes a review of the

Decision within that time.

Dated at Rockville, Maryland, this 6th day

of December , 1995.

For the Nuclear Regulatory Commission.

William T. Russell,

Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 95–30176 Filed 12–11–95; 8:45 am]

BILLING CODE 7590–01–P–M

[Docket Nos. 50–325 AND 50–324]

Carolina Power & Light Company;

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The U.S. Nuclear Regulatory

Commission (the Commission or NRC)

is considering issuance of an

amendment to Facility Operating

License Nos. DPR–71 and DPR–62

issued to the Carolina Power & Light

Company (the licensee) for operation of

the Brunswick Steam Electric Plant,

Units 1 and 2 (BSEP) located in

Southport, North Carolina.

Effective October 26, 1995, the

Commission amended its regulations

(10 CFR Part 50, Appendix J) to provide

a performance-based option for leakagerate testing of containments of lightwater-cooled nuclear plants. The

proposed amendment would permit the

licensee to implement this performancebased option, which allows leakage

testing intervals to be based on system

and component testing performance.

The proposed amendment requires

the establishment of a ‘‘Primary

Containment Leakage Rate Testing

Program’’ (program) and makes general

reference to the NRC guidance utilized

by the licensee for development of this

program, i.e. Regulatory Guide 1.163,

‘‘Performance-Based Containment LeakTest Program’’. Regulatory Guide 1.163

addresses the acceptability of industry-
63740 Federal Register / Vol. 60, No. 238 / Tuesday, December 12, 1995 / Notices

developed guidance described in

Nuclear Energy Institute document NEI

94–01, entitled ‘‘Industry Guideline for

Implementing Performance-Based

Option of 10 CFR Part 50, Appendix J.’’

The proposed amendment takes one

exception to the guidance in NEI 94–01.

Based upon the use of compensatory

measures, the exception would allow

the use of less accurate flow measuring

equipment.

Certain containment leakage testing

schedules and details regarding the

scope of containment valves and

penetrations to be leak-tested will be

included in the licensee’s program but

would be removed by this proposed

amendment from the BSEP Technical

Specifications. Consistent with NEI 94–

01 the proposed amendment relaxes the

schedules for performing primary

containment air lock leakage

surveillance testing and, if the interval

for testing of overall containment

leakage (Type A testing) has been

extended under the program to 10 years,

requires inspections for containment

integrity during two other refueling

outages before the next Type A test as

well as immediately prior to that test.

Before issuance of the proposed

license amendments, 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

amendments 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 license amendments do

not involve a significant increase in the

probability or consequences of an accident

previously evaluated. The proposed license

amendments revise the Technical

Specifications to reflect the adoption of a

performance-based containment leakagetesting program. The Nuclear Regulatory

Commission has approved the use of a

performance-based option for containment

leakage testing programs when it amended 10

CFR Part 50, Appendix J (60 FR 49495).

For adoption of the revised regulations,

licensees are required to incorporate into

their Technical Specifications, by general

reference, the NRC regulatory guide or other

plant-specific implementing document [used

to develop the performance-based leakagetesting program]. A new Administrative

Control subsection is being added to the

Brunswick Plant Technical Specifications

that requires the establishment and

maintenance of a Primary Containment

Leakage Rate Testing Program. As stated in

the Technical Specification, this Primary

Containment Leakage Rate Testing Program

will conform with NRC Regulatory Guide

1.163, Revision 0, dated September 1995,

‘‘Performance-Based Containment Leak-Rate

Testing Program’’ by establishing leakage

testing intervals based on the criteria in

Section 11.0 of NEI 94–01. The Technical

Specifications will continue to require

performance of a periodic general visual

inspection of the containment to ensure early

detection of any structural deterioration of

the containment system that might occur.

The effect of increasing containment

leakage rate testing intervals has been

evaluated by the Nuclear Energy Institute

using the methodology described in NUREG–

1493 [‘‘Performance-Based Containment

Leak-Test Program’’, September 1995] and

historical representative industry leakage rate

testing data. The results of this evaluation, as

published in NEI 94–01, Revision 0, are that

the increased risk corresponding to the

extended test interval is small (less than 0.1

percent of total risk) and compares well to

the guidance of the NRC’s safety goal.

Therefore, adoption of performance-based

verification of leakage rates for isolation

valves, containment penetrations, and the

overall containment boundary will provide

an equivalent level of safety and does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed license amendments will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated. No safety-related

equipment, safety function, or plant

operations will be altered as a result of the

proposed license amendment. The safety

objective for the primary containment is

stated in 10 CFR 50, Appendix A, ‘‘General

Design Criteria for Nuclear Power Plants.’’

The safety function of the primary

containment will be met since the

containment will continue to provide ‘‘an

essentially leak-tight barrier against the

uncontrolled release of radioactivity to the

environment * * *’’ for postulated

accidents. Therefore, the proposed license

amendments will not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

3. The proposed license amendments do

not involve a significant reduction in a

margin of safety. As stated above, the Nuclear

Regulatory Commission has approved the use

of a performance-based option for

containment leakage testing programs when

it amended 10 CFR Part 50, Appendix J (60

FR 49495). The new Primary Containment

Leakage Rate Testing Program will conform

with NRC Regulatory Guide 1.163, Revision

0, dated September 1995, ‘‘PerformanceBased Containment Leak-Rate Testing

Program’’ by requiring that leakage testing

intervals be established based on the criteria

in Section 11.0 of NEI 94–01, Revision 0.

As discussed in Part 1 above, the effect of

increasing containment leakage rate testing

intervals has been evaluated by the Nuclear

Energy Institute using the methodology

described in NUREG–1493 and historical

representative industry leakage rate testing

data. The results of this evaluation, as

published in NEI 94–01, Revision 0, are that

the increased safety risk corresponding to the

extended test intervals is small (less than 0.1

percent of total risk) and compares well to

the guidance of the NRC’s safety goal. In

addition, as demonstrated by risk analyses

contained in NUREG–1482 (sic) [NUREG–

1493], relaxation of the integrated leak rate

test frequency does not significantly increase

the probability or consequences of a

previously evaluated accident. Integrated

leakage rate tests have been demonstrated to

be of limited value in detecting significant

leakages from penetrations and isolation

valves. Therefore, the proposed license

amendments adopting a performance-based

approach for verification of leakage rates for

isolation valves, containment penetrations,

and the containment overall will continue to

meet the regulatory goal of providing an

essentially leak-tight containment boundary,

will provide an equivalent level of safety,

and do not involve a significant reduction in

a margin of safety.

The revised Technical Specifications will

continue to maintain the allowable leak rate

(La) as the Type A test [containment overall

leak-rate test] performance criterion. In

addition, a requirement to perform a periodic

general visual inspection of the containment

has been maintained as part of the

performance-based leakage testing program.

The revised Technical Specifications will

continue to maintain the allowable leak rate

(La) (sic) [0.6 La] as the Type B [containment

penetration leak-rate test] and C

[containment isolation valve leak-rate test]

tests’ performance criterion. As supported by

the findings of NUREG–1493, the percentage

of leakages detected only by integrated leak

rate tests is small (only a few percent) and

Type B and C leakage tests are capable of

detecting more than 97 percent of

containment leakages and virtually all such

leakages are identified by local leak rate tests

(LLRTs) of containment isolation valves.

Thus, the proposed license amendments do

not involve a significant reduction in a

margin of safety and will continue to ensure

the revised Appendix J regulatory goal of

ensuring an essentially leak-tight

containment boundary.

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.

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
Federal Register / Vol. 60, No. 238 / Tuesday, December 12, 1995 / Notices 63741

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. Should

the Commission take this action, it will

publish in the Federal Register a notice

of issuance and provide for opportunity

for a hearing after issuance. The

Commission expects that the need to

take this action will occur very

infrequently.

Written comments may be submitted

by mail to the Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, and should cite

the publication date and page number of

this Federal Register notice. Written

comments may also be delivered to

Room 6D22, Two White Flint North,

11545 Rockville Pike, Rockville,

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

Federal workdays. Copies of written

comments received may be examined at

the NRC Public Document Room, the

Gelman Building, 2120 L Street, NW.,

Washington, DC.

The filing of requests for hearing and

petitions for leave to intervene is

discussed below.

By January 11, 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 located at the

University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297. 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, 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 David

B. Matthews, petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, and to General Counsel,

Carolina Power & Light Company, P.O.

Box 1551, Raleigh, North Carolina

27602, attorney for the licensee.
63742 Federal Register / Vol. 60, No. 238 / Tuesday, December 12, 1995 / Notices

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 September 13, 1995,

as amended on November 27, 1995,

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 located at

the University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297.

Dated at Rockville, Maryland, this 5th day

of December 1995.

For the Nuclear Regulatory Commission.

David C. Trimble,

Project Manager, Project Directorate II–1,

Division of Reactor Projects—I/II, Office of

Nuclear Reactor Regulation.

[FR Doc. 95–30175 Filed 12–11–95; 8:45 am]

BILLING CODE 7590–01–P

[IA 95–058]

Five Star Products, Inc. and

Construction Products Research,

Fairfield, CT and H. Nash Babcock,

Order

I

Five Star Products, Inc. (FSP), is a

company located in Fairfield,

Connecticut, and was formerly known

as U.S. Grout Corporation. FSP

manufactures and sells grout and

concrete products to the nuclear

industry and has done so for about 20

years. Through a holding company, Mr.

Babcock owns FSP and several related

businesses, including Construction

Products Research, Inc. (CPR), which

performs laboratory tests of FSP

products. Mr. Babcock is Vice-President

of FSP and President of CPR.

II

FSP submitted its grout and concrete

products to CPR for testing. Following

the tests, CPR issued certifications that

it tested FSP products in conformance

with certain specifications of the

American Society for Testing and

Materials. FSP subsequently utilized

those certifications as the basis for

certifying that its products satisfied

Appendix B and customer Purchase

Order (PO) requirements. At various

times since 1980, FSP has advertised

and represented to NRC licensees that

its products are manufactured in

accordance with the requirements of

Appendix B. It has supplied products

pursuant to purchase orders requiring

FSP to meet the requirements of

Appendix B, and 10 CFR Part 21.

Licensees who have purchased material

from FSP under FSP’s certification of

quality have used the grout and concrete

in safety-related applications and as

basic components.

The Nuclear Regulatory Commission

(NRC or Commission) issued 10 CFR

Part 21 (Part 21) to implement Section

206 of the Energy Reorganization Act of

1974. Part 21 imposes, inter alia,

evaluation and reporting requirements

on directors and responsible officers of

firms which supply basic components of

any facility or activity which is licensed

or otherwise regulated pursuant to the

Atomic Energy Act of 1954, as amended,

or the Energy Reorganization Act of

1974. Basic components are structures,

systems, or parts in which a defect or

failure to comply with applicable

requirements could create a substantial

safety hazard. 10 CFR 21.3(a). Part 21 is

implemented in conjunction with

Appendix B, which contains the quality

assurance (QA) criteria applicable to

design, fabrication, construction, and

testing of safety-related structures,

systems, and components in commercial

nuclear power plants. Together, these

requirements are intended to assure the

safety of safety-related components,

materials, and services for nuclear

power plants.

Section 206 of the Energy

Reorganization Act of 1974 requires

directors and responsible officers of

firms constructing, owning, operating or

supplying the basic components of a

facility or activity licensed or regulated

by the Atomic Energy Act of 1954, as

amended, who obtain information

regarding defects in those basic

components, or failures of basic

components, or of the facility to comply

with NRC requirements, to notify the

NRC of those defects and failures to

comply. Section 206(d) authorizes the

Commission to conduct inspections and

other enforcement activities necessary

to insure compliance with that section.

10 CFR 21.41 and 21.51 implement

Section 206(d).

III

The NRC conducts inspections of

vendors who supply safety-related

components pursuant to Appendix B

and who supply basic components

pursuant to Part 21. On August 18,

1992, the NRC began an unannounced

inspection of FSP, and of its laboratory

contractor, CPR, to determine the extent

to which FSP supplied basic

components to NRC licensees, the

adequacy of FSP’s QA Program, the

adequacy of CPR’s testing of FSP

products, and the adequacy of FSP

products.

Shortly after the inspection began, Mr.

Babcock met with the inspection team

and questioned the NRC’s authority to

conduct the inspection. Mr. Babcock

was presented with two identical letters

from the NRC staff, dated August 13,

1992, each addressed separately to FSP

and CPR. The letters outlined the NRC’s

inspection authority under 10 CFR Part

21, Section 161o of the Atomic Energy

Act of 1954, as amended (AEA), and

Section 206(d) of the Energy

Reorganization Act of 1974, as amended

(ERA). Despite this, Mr. Babcock

continued to question the NRC’s

authority and, throughout the

inspection, denied the inspectors access

to inspect CPR’s testing laboratory,

which was located in the basement of

FSP’s Fairfield, Connecticut,

headquarters, and access to inspect

CPR’s laboratory records.

During the inspection of August 18

and 19, 1992, the inspection team

reviewed NRC power reactor licensee

POs submitted to Five Star in order to

determine the scope of FSP’s nuclear

involvement. The team was provided

with POs for the period 1988 to 1992.

Those POs demonstrate that at least

seven NRC reactor licensees and one

licensee contractor had issued POs to

FSP for safety-related grout and concrete

mix products, and had specified

compliance with Appendix B and Part

21.

The inspection team reviewed copies

of several NRC licensee audit reports of

FSP and CPR. These reports

documented that NRC licensee requests

to audit CPR’s test laboratory and

records were consistently denied by

FSP. Further, several NRC licensee audit

reports found that FSP’s QA program

was not acceptable and did not meet

certain requirements of Appendix B.

The NRC inspection team requested

copies of all audits performed by FSP of

CPR to determine CPR’s compliance

with the quality assurance criteria of

Appendix B and Part 21. Only one FSP

audit of CPR was performed, by the FSP

QA Manager, and it was provided to the

NRC inspection team by the FSP QA

Manager. The July 31, 1992 audit report

concluded that CPR’s June 10, 1992 QA

program was satisfactory. The format

and most of the language of this report

were identical to a report of an audit

conducted by Toledo Edison, an NRC