ML20151V451

Insp Repts 50-313/88-04 & 50-368/88-04 on 880229-0310. Violations Noted.Major Areas Inspected:Supplemental Design Control,Allegation assessment,10CFR21 Insp,Inservice Insp & Design Changes & Mods

ML20151V451
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/13/1988
From:	Barnes I, Gilbert L, Renee Taylor NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20151V435	List: ML20151V432 ML20151V438 ML20151V451
References
50-313-88-04, 50-313-88-4, 50-368-88-04, 50-368-88-4, NUDOCS 8805030064
Download: ML20151V451 (17)

See also: IR 05000313/1988004

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APPENDIX B

U.S. NUCLEAR REGULATORY COMMISSION .

REGION IV

NRC Inspection (eport: 50-313/88-04 Operating Licenses: OPR-51

50-368/88-04 NPF-6

Dockets: 50-313  !'

50-368

Licensee: Arkansas Power & Light Company (AP&L)

Facility Name: Arkansas Nuclear One (ANO), Units 1 and 2

Inspection At: ANO, Russellville, Arkansas

Inspection Conducted: February 29 through March 10, 1988

Inspectors: /5/TI

. D. Gilbert, Reactor Inspector, Materials Date ' i

and Quality Programs Section, Division of .

4 Reactor Safety

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/_ bT/ d) G

R. 'G. Taylor'/ Reactor Inspector, Materials '

andQualittProgramsSection,Divisionof D.dtf l

Reactor Safety

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l 8 n,a- 4t.//3 h y

I. Barnes, Chief, Materials and Quality Date

Programs Section, Division of Reactor Safety

c Accompanying

! Personnel: R. C. Haag, Materials and Quality Programs Section, Division of l

Reactor Safety '

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C. D. Sellers, Materials Engineering Branch, Office of Nuclear

Reartor Regulation

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Approved: I 3 n- ~ , +-/cr/r,f'

I. Barnes, Chief, Materials and Quality Date  ;

Programs Section, Division of Reactor Safety

5805030064 880422

PDR ADOCK 05000313

O DCD

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Inspection Summary

Inspection Conducted February 29 through March 10, 1988 (Report 50-313/88-04)

Areas Inspected: Routine, unannounced inspection of supplemental design

control, allegation assessment, and 10 CFR Part 21 inspection.

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Results: Within the three areas inspected, one violation was identified

(f aiTure to enter potential 10 CFR Part 21 items into the required tracking

system, paragraph 7). Two unresolved items are identified in paragraphs 4 and

7.

Inspection Conducted February 29 through March 10, 1988 (Report 50-368/88-04)

Areas Inspected: Routine, unannounced inspection of inservice inspection,

design changes and modifications, supplemental design control inspection, use

of E-Brite 26-1 tubing in shutdown cooling heat exchangers, allegation

assessment, and 10 CFR Part 21 inspection.

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Results: Within the six areas inspected, two violations were identified

(failure to properly document ultrasonic examination calibrations, paragraph 2;

and failure to enter potential 10 CFR Part 21 items into the required tracking

system, paragraph 7). Two unresolved items are identified in paragraphs 3 and

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DETAILS t

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l 1. Persons Contacted

l AP&L

l + 5. M. Quennoz, Plant General Manager

+ L. Humphrey, General Manager, Nuclear Quality

+*R. Lane, Manager, Engineering '

• J. McWilliams, Manager, Maintenance

• D. Howard, Manager, Licensing

+*J. L. Taylor-Brown, Quality Control / Quality Engineering (QC/QE)

Superintendent

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• B. A. Baker, Modifications Manager

+*D B. Lomax, Licensing Supervisor

+ R. A. Courtney, Quality Engineering / Nondestructive Examination (QE/NDE), '

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AP&L Consultant

1 + G. D. Provencher, Quality Assurance (QA) Supervisor

+ A. B. McGregor, Engineering Services Superintendent

D. Payne, Inservice Inspection Coordinator

, M. Smith, Supervisor, Nuclear Engineering

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O. Spond, Metallurgist

• M. Snow, Licensing Engineer l

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Combustion Engineering (CE) I

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i D. D. Weber, Principle Field Service Engineer i

R. Kusy, Examiner (UT) Level III

NRC

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• W. D. Johnson, Senior Resident Inspector

• C. C. Harbuck, Resident Inspector

l The NRC inspectors also interviewed other licensee and contractor

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employees during the inspection.

• Denotes those attending the exit interview on March 4, 1988.

j + Denotes those attending the exit interview on March 10, 1988. ,

j 2. Inservice Inspection (ISI) (73051, 73052, 73753, and 73755)

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a. Program Review

The NRC inspector reviewed AP&L Procedure No. 1092.25, the ISI l

, Administrative Procedure, and the ISI Program Plan and Schedule for l

the sixth refueling outage for Unit 2. In reviewing the changes

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which had been made to the ISI Program Plan for this outage, the NRC

j inspector noted that several examination numbers for zones 1 and 2

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did not coincide with the numbers on the drawing. The licensee

issued a change to the Program Plan to correct the discrepancies

which resolved the NRC inspector's concern.

During revicw of the ISI Program Plan, the NRC inspector requested

the repair history on the reactor vessel beltlins region material,

The licensee indicated that the weld repair history would be obtained

from the manufacturer of the reactor vessel and any repaired areas

requiring subsequent examination would be incorporated into the ISI

Program Plan. ASME Code,Section XI of the 1974 Editior. through

Summer 1975 addenda requires examination of certain repaired areas

when the repaired areas have been exposed to neutron flux in excess

of 10" nyt. Since the neutron flux was determined to be 9.332 x

10" nyt at this outage for Unit 2 and even less for Unit 1, the

examination of repaired areas is not required at this outage. This

is an open item pending the licensee's determination ar, to whether

repairs were made to the beltline region material and the

incorporation of any required examinations into the ISI Program Plan

for both Units 1 and 2. (313/8804-01; 368/8804-01)

b. Review of procedures

The NRC inspector reviewed the procedure for performing the automated

ultrasonic examination of the reactor pressure vessel welds; i.e.,

AP&L Procedure No. 1092.021, Revision 0, Attachment I - ANO-410-002,

Revision 1.

In the areas reviewed, the procedure was consistent with the

requirement of ASME Section XI, Regulatory Guide (RG) 1.150 and Code

Case N234,

c. Observation of Work Activities

The NRC inspector observed the automated ultrasonic examinatior. of

Exam No.01-012 for the lower shell to middle shell circumferential

weld and Exam No.01-014 for the middle shell longitudinal weld at

210'. The examinations were consistent with the Program Plan for

method and extent of examination. The personnel performing the

examinations were certified as required by Procedure ANO-410-002 for

examiners and Procedure 151-046 for ISI-2 equipment operators. The

certification documents for the eight search units used on Scan

No. 01-012-1 with Orientation 2 for Exam No.01-012 were reviewed and

found to be consistent with the angles and frequencies required by

Procedure ANO-410-002.

d. Data Review and Evaluation l

The NRC inspector reviewed the records for Data Package j

No. ANO-410-002-4 for Exam No.01-017. This examinetton had been l

completed and the examination data was in the process of being 1

evaluated. Exam No.01-017 was an automated ultrasonic examination

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of the reactor vessel upper shell longitudinal seam weld at a i

location of 90'. The examination was performed from the inside of

the reactor vessel using a sled assembly containing the five search l

units used to conduct the ultrasonic examination at a O' angle, two

45' angles, and two 60' angles.. During review of the calibration  ;

records for the five search units, the NRC inspector noted that the  !

calibration sheet for the 60' angle search units was unsigned and the !

two calibration sheets for the 45' angles and the O' angle search l

units were initialed by a Level II examiner but not dated. It was ,

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also noted that two of the five calibration checks had not been

signed on the examination record for Exam No.01-017, and one of

these was for verification that the initial calibration had been

established. j

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Paragraph 8.0 and Appendix 0 of Procedure 1092.021, Attachment I,  !

require that the Level II examiner shall establish the system l

calibration parameters and sign and date the calibration sheet prior 1

to the ultrasonic examination of the reactor vessel welds. The

procedure also requires that the calibration checks will be

documented by time, date, and signature.

This failure to follow procedural requirements for documenting I

calibration activities is an apparent violation. (368/8804-02)

3. Design Changes and Modifications (37700)  !

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An inspection was conducted to ascertain whether design changes and i

modifications, which had been determined by the licessee as not requit?ng l

NRC approval, were in conformance with the requirements of the Technical  !

Specifications and 10 CFR 50.59. Modifications associated with the ,

current Unit 2 outage (2R6) were reviewed.  ;

The NRC inspector reviewed the modifications itsted below for which work i

was in progress. Each modification involved a design change to plant as  !

described in the licensing basis documents (LB0s). This allowed  !

observation of the licensee's program for controlling work that affects i

existing plant design. The design change packages (DCPs) for the '

modifications were also reviewed.

85-2174, "Flow Element for Service Water Pump Test"

• 86-2100, "LPSI System Overpressurization Protection"

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86-2110 "Replace Service Water Pipe to HPSI Pumps with SS"

During this review, the NRC inspector focused on the two onsite

engineering groups (Project Engineering and Field Engineering) that

directly deal with modification implementation. Project Engineering

receives DCPs from the licensee's design engineering group located in

Little Rock, Arkansas, and coordinates the installation effort. Once the

DCP is received onsite, changes to the package are performed by project

engineering via a DCP revision. Field engineering receives the DCPs and

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specific installation requirements from Project Engineering, and from this

information they formulate Construction Work Permits (CWPs), which provide

detailed work instructions.

In DCPs 85-2174 and 86-2110 modifications are performed on both trains A

and B of the service water system; however, during the outage, one train

must remain operable. The NRC inspector reviewed the "Installation Plan

Review and Approval," which describes the overall scope of the job and

lists the sequence of installation attributes required to ensure one train

operability while the modification is being accomplished. This plan also.

addresses 10 CFR Part 50.59 considerations for the time period the

modification is taking place. Project engineering, via the installation

plan, appeared to be properly coordinating modifications and prescribing

requirements for system turnover following modification.

During the review of DCP 85-2174, the NRC inspector queutoned the bisis

of calculation No. 88E00011-01._ This calculation dealt with piping

supports that are common to both trains of the service water system.

During various times of the modification, three "common" supports will

become disabled. This calculation was to verify one of these supports

could be disabled and still maintain system operability. The rationale of

the calculation involved a review of the piping stress analysis with all

the existing supports in place, then concluding by "engineering judgement"

that one of the three supports could be declared inoperable. This

engineering judgement was based on "the many operability studies made for

other nuclear plants where removal of one support between anchors meet the

plant operability limits." The lack of details (i.e., piping stresses,

support loadings) in correlating this situation with the "operability

studies made for other nuclear plants" caused the NRC inspector to

question the validity of this calculation. This subject will remain an

unresolved item pending the licensee providing a more detailed

justification for the comparison used in the calculation and NRC review of

the additional data. (068/8804-03)

The 10 CFR Part 50,59 reviews for the three DCP were technically adequate

in providing a bases for the unreviewed safety question determination.

The NRC inspector also reviewed the ANO system for changing operating

procedures and updating operator training to reflect plant modifications.

The system is comprehensive in addressing all modifications initially,

then using the system turnover process to verify required training and

procedure changes are performed. For the DCPs that were being worked, the

NRC inspector verified that the applicable procedure changes and operator

training were in the final state of being incorporated.

The NRC inspector reviewed the process for identifying upcoming

modifications on plant drawing, i.e., P& ids, isometrics, and hanger

details. During DCP development, drawings associated with the

modification are changed. The drawing is then assigned a numerical

designator (dash number) after the revision to uniquely identify the

change on the drawing to a particular DCP. For two of the drawings

reviewed by the NRC inspector, five separate dash numbers had been

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reserved. The information from previous dash numbers is not shown on

l changes to drawings that deal with a different DCP and new dash number. *

Subsequent changes to a modification and particular drawing dash number

are made by a DCP revision. While no specific problems were found while

reviewing drawings, the NRC inspector noted that this process relies on

the project engineer's knowledge of his job and other related '

modifications when making changes to drawings with dash numbers.

No violations or deviations were identified.

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4. Supplemental Design Control Inspection

The purpose of this phase of the inspection was to:

a. Obtain additional information relative to a modification of Unit 3

that should have occurred in 1985-1986 as a result of followup on a i

problem which developed during a Unit 2 containment integrated leak '

rate test performed in 1985.

b. Obtain additional information relative to what corrective actions the

licensee may have taken as a result of a above.

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In regard to Item e, above, the NRC inspector had learned that, during the

integrated leak rate test of Unit 2 containment in 1985, the licensee had )

identified a leakage path in Post-Accident Sampling System (PASS) and had

taken prompt eagineering action to rectify the problem in Unit 2 in May ,

1985. The leakage path identified was created when two isolation valves 1

were installed such that they isolated the PASS flow to the containment

but did not isolate flow from the containment to the PASS or to the '

atmospheric environment. A similar installation had been accomplished in f

the 1980-1981 timeframe in Unit 1 in the same manner; i.e., the PASS

isolation valve was installed backward. When the leakage problem was

identified in Unit 2, an examination of the Unit I design was performed

and a correction was developed as DCP 85-1036 and was approved for

installation in April 1986. The package was scheduled for installation

during the refueling outage of 1986 which took place from September

through December. The NRC inspector was unable to specifically establish

why the change was not accomplished during the outage and did not occur  !

until February 1988 with Unit I at near full power. This subject is I

considered unresolved pending NRC issue of the results of the Unit 2

integrated leak rate testing inspection. (313/8804-02)

In addition to simply reversing the isolation valve, the DCP 85-1036 also

incorporated a TP connection to allow for Type C testing of the

installation. As a collateral matter, the NRC inspector also reviewed QC

records covering the modification and test of the modification. Tne

records indicated that QC personnel had performed inspactions of the basic

modification work such as welding and nondestructive examinations and, in

addition, observed local leak rate testing of the modification after the

change was complete.

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, In regard to Item b. above, the NRC inspector was informed that th  !

] licensee had developed an improved program for centrol of both the '

engineering contents of DCPs and a prioritization control system to better

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ensure incorporation of changes in a timely manner. The NRC inspector  ;

reviewed Procedure 6010.001, "0CP Development," dated September 17, 1987,  :

i which details the controls placed on package engineering content. Time t

. did not allow for an evaluation of the implementation of this procedure  !

! w;iich could only be accomplished by detailed review of several recent j

packages. The NRC inspector also reviewed Procedure 6000.40,"Project  ;

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Planning and Control," which was issued on July 15, 1987. This document l

outlines all of the management, budget, and scheduling controls the

licensee applies to design changes. The document also requires the person j

i or the group developing a change package to indicate a priority for

accomplishing the modification work. This priority does not appear to be

based on either time or on milestone events for the affected unit but  !

j rather only on high, medium, or low urgency. The licensee also established j

a site priority review committee. Individual members vote to confirm a ,

priority assignment or to change the assignment. The votes are recorded  :

I in. committee minutes. This same committee is responsible for reviewing i

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the outage modification schedule developed by others and confirming or ,

i recommending changes to the schedule. The Executive Director of Nuclear

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Operation (plant manager) is the final approving authority on all design  ;

j  ?.snges and schedules for accomplishment thereof. Implementation of  ;

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current design change controls is considered an open item pending i

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additional NRC inspection. (313/8804-03; 368/8804-04). l

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5. Use of E-Brite 26-1 Tubing in ANO Unit 2 Shutdown Cooling Heat Exchangers l

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4 a. Background [

)! In June 1986, the NRC Region IV office was notified of an allegation l

contained in a June 11, 1986, letter to President Reagan concerning a  !

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high purity ferritic stainless steel product identified as '

E-Brite 26-1. A file of copies of material test reports generated by  !

I or for Air Reduction Company (AIRCO), covering a time period from  !

i 1970 through 1976, was subsequently furnished by the alleger to the

1 NRC Region IV office. The letter and furnished information were t

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forwarded to the NRC Vendor program Branch (VPB) for followup.  !

i During an inspection in August 1986, which is documented in NRC 1

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Inspection Report 99901063/86-01, a VpB inspector met with the  !

alleger to gather information on the allegation. From this meeting

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it was ascertained that the alleger apparently believes that '

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E-Brite 26-1 material cannot be welded and its ductile to brittle

i transition temperature cannot be controlled, thus making its use in ,

j nuclear reactors extremely hazardous. The information provided by the i

! alleger pertained to E-Brite 26-1 material r.anufactured by AIRCO

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Vacuum Metals in Berkeley, California. The alleger believed that i

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E-Brite 26-1 material is p m ent in a number of domestic nuclear

j power plants. The results d the VPB inspection did not substantiate

j the allegation concerning use of E-Brite 26-1 materials in domestic

nuclear power plants.

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In late 1987, the NRC staff ascertained that two shutdown cooling

heat exchangers at ANO Unit 2 had been retubed in 1981 with

E-Brite 26-1 material furnished by the current patent holder for the

material, Allegheny Ludium Steel Corporation. A followup inspection

has been performed by the VPB at the tubing manufacturer's facility.

The purpose of this inspection was to review the procurement and

operational history associated with retubing of the ANO Unit 2

shutdown cooling heat exchanger;, with E-Brite 26-1 material.

b. Procurement of Replacement Tube Bundles

The NRC inspectors discussed with the AP&L metallurgist the

background relative to retubing of the ANO Unit 2 shutdown cooling

heat exchangers with E-Brite 26-1 material. The AP&L metallurgist >

informed the NRC inspectors that the final selection of E-Brite 26-1

was based on its excellent corrosion properties and assurance r,f

adequate weldability resulting from discussions with Allegheny

l Ludium. The selected fabricator for the replacement tube bunales,

l Engineers and Fabricators Co. (EFCO), was an ASME certificate holder

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for "N" and "NPT" stamps. EFC0 had prior experience in welding

E-Brite 26-1 for nonnuclear applications and did not anticipate any

problems in making tube-to-tube sheet welds.

(1) Procurement Documents

The NRC inspectors reviewed the technical and quality

requirements contained in Purchase Requisition (PR) 11264, which

was applicable to the procurement from EFC0 of the two

replacement tube bundles (Equipment Nos. 2E35A, 2E358).

Included in the PR was Combustion Engineering Specification

, No. 6730-PE-301, "Project Specification For A Shutdown Heat

l Exchanger," whi &. had been revised to specify E-Brite 26-1 as

the tubing material in accordance with the requirements of ASME

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Section III Code, Class 2, 1980 Edition. The applicable ASME

material specification for this alloy steel tubing is SA-268

Grade XM 27. The stipulated tube side design pressure and

temperature were, respectively, 650 psig and 400'F. The

replacement tube bundles were required to meet the requirements

of ASME Section XI Code, 1980 Edition, Article IWC-7000.

l (2) Allegheny Ludlum Tubing Manufacture

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The NRC inspector reviewed Allegheny Ludlum Tubular Products

Company Quality Assurance Policy and Procedure No. 0202,

Revision 0, effective date December 1,1980, "Quality System

Program." Also reviewed were the EFC0 report of the quality

assurance survey performed at Allegheny Ludlum and flow diagrams

showing applicable manufacturing operations, inspection 0 and

tests. From discussions with the AP&L metallurgist, it was

ascertained that he had visited the Allegheny Ludlum facility to

witness in process fabrication of the E-Brite 26-1 welded tubing

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for the ANO Unit 2 shutdown cooling heat exchangers. The trip  !

l report for this activity was reviewed, from which the following i

, pertinent information was learned. Allegheny Ludlum was

l cognizant of the need to closely control the welding process in l

order to minimize both weld defect occurrence and elevated '

ductile to brittle transition temperatures in the weld zone. As

an in process control, a reverse bend test was performed once i

per hour on a welded section at a temperature of 32'F. Liquid

penetration examination was also utilized for in process control

of welding, with the first five tubes from each skelp roll .

inspected and then every subsequent fifth tube. Further i

assurance of weld joint integrity was provided by eddy current I

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testing (in accordance with the requirements of ASME

Specification SA-655), pneumatic testing at 150 psig air

pressure with the tube submerged in water, and hydrostatic ,

testing of finished tubes at 1000 psig.

The NRC inspectors reviewed the Allegheny Ludlum certified

material test reports (CMTRs) for the furnished tubes and

determined that the chemical analysis and mechanical test  !

results were in compliance with the requirements of SA-268,

Grade XM 27. During this review, one of the NRC inspectors also  !

verified that bright annealing (followed by forced cooling) was  !

documented as having been performed. This verification was  :

performed both as a result of the identification to the AP&L

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metallurgist by Allegheny Ludium that they possessed data which l

shows the annealing cycle reduces the ductile to brittle ,

transition temperature, and to provide assurance that the  !

material was not held at temperatures which embrittle high  ;

chromium ferritic alloys; i.e., 700-1100*F. i

(3) EFC0 Tube-to-Tube Sheet Welds

The NRC inspectors reviewed the following EFC0 documents which i

were considered pertinent to tube-to-tube sheet welding of the

replacement heat exchanger bundles

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Welding Procedure Specification (WPS) No. 015316L, dated

January 1981, for clad plate; i.e., shielded metal arc weld '

overlay of tube sheets;

WPS No. T101.316LTO, dated January 1981, for gas tungsten

are tube-to-tube sheet welds; i.e., SA-268 Grade XM 27 l

tubes to Type 316L overlay cladding; and ,

• supporting procedure qualification records for the WPSs.

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No problems or technical discrepancies were identified during l

this review with respect to the requirements of Section III and

Section IX of the ASME Code. The NRC inspectors also examined a

section which had been retained of the tube-to-tube sheet

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qualification welds. This sample showed fully acceptable welds

with no evidence of defects. The NRC inspectors also reviewed

applicable NDE procedures (i.e., Specification VO-1, Revision 3,

"Ultrasonic Inspection of Bond of Weld Overlay to Base

Material," and Specification NLP-2, Revision C, "Liquid

Penetrant Inspection") and AP&L trip reports for surveillance

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activities. From the review of the trip reports, the NRC *

inspectors noted that some problems were encountered in

performing the 2E358 bundle tube-to-tube sheet welds.

Approximately 16 welds were found to leak when tested under

100 psi air pressure and required repair. The AP&L metallurgist

considered that the most probable cause of these leaks was

incomplete removal of soap solution from tubes prior to

tube-to-tube sheet welding. The soap solution was applied as a

lubricant to the tubes during bundle asserrbly operations. All

welds were examined by the liquid penetrant method after welding

and also after repairs. Also neted from the trip reports was

that tubing damage (which required replacement of some tubes)

occurred during insertion of the "B" bundle into the shipping

shell,

c. Operational History

The NRC inspectors were informed that shutdown cooling heat

' exchanger 2E35B had operated without any problems since installation

of the replacement tube bundle. (The 2E35A replacement tube bundle,

which had been fabricated without problems, was installed in the

2E358 shutdown cooling heat exchangers as a result of delivery

schedule considerations.) Initial leakage problems were, however,

encountered with the 2E35B replacement tube bundle after installation

in shutdown cooling heat exchanger 2E35A. A total of 18 tubes were

plugged in January 1983 as a result of testing using multiple leak

detection methods. No further leakage problems were stated to have

occurred. One of the NRC inspectors confirmed the accuracy of the

above information by review of a computer printout of all issued Job

Orders and Reports of Abnormal Occurrence that were applicable to the

shutdown cooling heat exchangers,

d. Summary of Findings '

The NRC inspectors determined from review of the Allegheny Ludium

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CMTRs that the shutdown cooling heat exchanger tubing was in

compliance with the chemical composition, mechanical properties, and l

nondestructive examination requirements of ASME Material l

Specification SA-268 Grade XM 27. This compliance, coupled with the

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in process controls used by the tubing manufacturer and also the

satisfactory hydrostatic test results obtained by both the tubing

manufacturer and replacement bundle f abricator, provide assurance of

the integrity of the tubing and absence of significant defects in the

weld joint region. The small section thickness of the tubing used

for the tube bundles (0.049 inches and 0.065 inches) would

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additionally be expected to result in ductile to brittle transition l

} temperatures that are at a minimum for the material in question. i

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Tube-to-tube sheet leakage problems were encountered during ,

fabrication and initial service of one of the two replacement tube l

i bundles. Since correction of this condition in January 1983, the

maintenance records confirm that the heat eschanger h o operated

3 satisfactorily without any further problems. Similarly, the  ;

i maintenance records confirm that the replacement tube bundle in the i

other heat exchanger has performed satisfactorily since installation, j

l The NRC inspectors concluded from review of the manufacturing,

examination, and test records associated with the manufacture of the

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j replacement tube bundles, that the Allegheny Ludlum E-Brite 26-1

material has adequate weldability for this application. The absence

of operational problems with the shutdown cooling heat exchangers  !

since January 1983 provides additional assurance that utilization of '

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this material for this application was an appropriate selection. l

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1 No vielations or deviations were identified in this area of the

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j 6. Assessment of Allegations .

. (Technically Closed) Allegation RIV-86-A-0075: This allegation

indicated that: (1) the position of Manager, Nuclear Quality Control was ,

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lowered to Superintendent, Quality Control; and (2) the minimum qualifications i'

of General Manager, Nuclear Quality were made less restrictive than had

i been the earlier position of Manager, Nuclear Quality Control. The

j' alleger provided the NRC with copies of three position descriptions. The ,

titles and dates of the position descriptions are as follows*  !

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j Manager of Nuclear Quality Control (position 0370.0) February 1982

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Superintendent, Quality Control (position 0370.0) February 1986

General Manager, Nuclear Quality (position 0483.0) February 1986

During an interview with the licensee's General Manager, Nuclear Quality,

, the NRC inspector observed that the position descriptions 0483.0 and

0370,0 dated February 1986 were in a folder in his possession and that

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they were the same as those offered by the alleger. The official purpose

i of the position descriptions was not clearly established although it was I

understood that they are largely used to establish salary ,br positions. j

, The General Manager agreed that the Superintendent of Quality Control was ,

of somewhat lower organizati nal stature than was the predecessor position l

of Manager, Nuclear Quality Control but also offered that the newer

a position of General Manager was intended to combine the authorities and

] responsibilities of the earlier positions of Manager, Nuclear Quality

Control and Quality Assurance Manager. The General Manager indicated that 1

i the purpose of the change was to give greater visibility to the Quality

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Assurance organization. The General Manager offered the NRC inspector a

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copy of the licensee's 1985 version of the Quality Assurance Manual

Operations (Revision 7) that was submitted to the NRC in the summer of

1985. This manual describes what amounted to two separate quality

organizations. One of the units was located in the corporate offices and

was headed by the Quality Assurance Manager. This unit was responsible

for audits and surveillance of site operations and performed inspections

of vendor activities. The Quality Assurance Manager reported to the

Director of Energy Supply Services. The Manager, Nuclear Quality Control

was located at the site and was responsible for conducting inspections of

site activities such as maintenance and modification work and he reported

to the General Manager of Plant Operations. Revision 8 of the QA Manual

as submitted to the NRC in the summer of 1986 described the newer

organization consisting of: the General Manager, Nuclear Quality; the

Superintendent, Quality Control; and a Superintendent, Quality Assurance.

The new organization combi.1ed all of the responsibilities of the two

predecessor organizations into one organization which reports to the

Nuclear Operations Vice President. Both organizational structures are

suitable under Criterion I of Appendix B to 10 CFR part 50 and both were

accepted by NRC personnel responsible for reviewing the licensee's QA

program submittals.

Allegation I is considered substantiated as stated since the new position

of Superintendent, Quality Control is of less stature than the predecessor

position of Manager, Nuclear Quality Control. The allegation is not,

however, considered to have safety or regulatory merit since all of the

functions of the former position were transferred to the new organization,

although distributed differently, and collectively report to higher level

of management authority than did either of the earlier organizations.

Relative to Allegation 2, the NRC inspector observed that Position

Description 0370.0 dated February 1982 required.that the Manager, Nuclear

Quality Control have a bachelor's degree in enginstring or applied

sciences and that he have a minimum of 5 yea >'s M experience in industrial

Quality Assurance of which 2 years must have h en in a nuclear power plant

operation. Position Descriptions 037E 0 and 0483.0 dated February 1986

both state that, as a minimum, the persons occupying the positions have a

bachelor of science degree in engineering or equivalent and 5 years

experience of QA/QC related work, of which 2 years should be nuclear QA

experience. (The underlined words above are those words underlined by the

alleger to emphasize his concern.)

Based upon the position descriptions, the NRC inspector concluded that

the allegation was substantiated. Having reached this conclusion, the NRC

inspector again interviewed the General Manager, Nuclear Quality. The

General Manager indict.ted that it was his belief that even though the

experience and educational minimums stated in the position descriptions

had been made less stringent, the change neither violated any commitments

to the NRC nor the Technical Specification requirements. He again offered

the two versions of the QA Manual Operations. Revision 7 in 1985 provided

mininum qualification requirements for the Quality Assurance Manager but

was silent on those applicable to the Manager, Nuclear Quality Control.

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Revision 8 of the manual in 1986 provides minimum qualification

requirements for the General Manager, Nuclear Quality. In Revision 7, the

Quality Assurance Manager was to have a degree from an accredited school

in engineering or a related scientific discipline or equivalent with

5 years of experience in QA or QC of which 2 years was to be in the

nuclear field. Revision 8 places the same educatiorial requirements on the

General Manager but reduces the experience requirement to 4 years with a

stipulation that 1 year be in supervisory QA position.

The Technical Specifications for ANO Unit I require compliance to

t ANSI N18.7-1971 in the area of personnel qualifications while ANS 3.1-1981

is specified for Unit 2. The newer requirement indicates that QA Manager

should have an engineering or science degree with 4 years of experience,

one of which sho>ld be in a supervisory QA position. ANS 3.1 also states

that no candidate should be excluded from consideration because he has not

obtained a degree and suggests ten factors that may be considered in lieu

thereof. One of the factors is experience which essentially doubles the

4 basic experience requirements. It is apparent that the "equivalent" in

the educational requirements in the licensee's stated requirements is

intended to encompass additional experience in lieu of education.

TheGeneralManageralsostatedduringtheinterviewthathehasbeen

granted a master s degree in industrial engineering by the University of

Arkansas and that his subordinate Superintendent, Quality Control has a

bachelor's degree in chemical engineering. The General Manager also far

surpasses the minimum experience requirements.

The NRC inspector concluded that while the allegation is technically

substantiated, it was without safety or regulatory significance since the

licensee's commitments as contained in the QA manual and in the Technical

q Specification have been consistently satisfied. Further, the present i

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incumbents of senior positions in the QA organization exceed the minimum

requirements by a substantial margin, i

No violations or deviations were identified in this area of inspection.

7. 10 CFR part 21 Inspection (36100) l

The purpose of this inspection was to determine whether the licensee had

established and implemented procedures and controls which provide for

j evaluating of deviations, assuring that defects or failures to comply are

t reported to the NRC and that records applicable to these activities are

established and maintained.

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a. Posting

The NRC inspector examined the main notice board at the ANO site and

i verified tnat a notice which complied with the posting requirements

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of 10 CFR Part 21 was present.

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! No violations or deviations were identified in this area of

inspection. -

b. Review of Procedures

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identified by the posted notice as one of the applicable

organizations for site based individuals to report potential 10 CFR

Part 21 concerns, that the governing procedure for 10 CFR Part 21 was

AP&L Energy Supply Procedure 304, "Resolution of Nuclear Safety or

Nuclear Environmental Concerns." From discussions with site

licensing personnel and review of Revision 2 of this procedure, the

NRC inspector determined that the primary responsibilities for

evaluation, tracking, and resolution of potential 10 CFR Part 21

items were assigned to Little Rock, Arkansas, staff. It was also I

ascertained that the current revision of this procedure (Revision 3)

had apparently not been forwarded from Little Rock to the ANO site

licensing organization.

ANO site licensing obtained a c.omplete "Possible Nuclear Safety or

Environmental Concern Log" for both ANO units from Little Rock for

the NRC inspector to review. As a result of review of these logs,

the NRC inspector noted thai final resolution of two potential 10 CFR *

Part 21 items which had been received, respectively, on November 2,

1983, and February 12, 1985, had not been accomplished. Also noted

was several potential 10 CFR Part 21 items which had been received in

i 1986, which vere still in an open status. AP&L Energy Supply

! Procedure 304, Revision 2, provided a mechanism for tracking (and

management awareness of status) of potential 10 CFR Part 21 concerns

in that the procedure required entry of these concerns into the

licensing commitment tracking system. The NRC inspector confirmed

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that paragraph 5.5 of Revision 3 of this procedure contained the same '

requirement. The NRC inspector ascertained, however, from review of

the licensing commitment tracking system that only one of the

curren*.ly open potential 10 CFR Part 21 items was identified. The

failure to implement procedural tracking commitments is an apparent

violation. (313/8804-04,368/8804-05)

In order to assess 10 CFR Part 21 procedural controls with respect to

the procurement process and interfaces with AP&L Energy Supply

Procedure 304, Revision 3, the NRC inspector reviewed the following

documents:

Administrative Procedure No. 1000.08, "NRC Reporting and

Communications," Revision 3, dated June 1,1987;

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AP&L Energy Supply Procedure 213, "Nuclear Procurement,"

Revision 0, dated March 20, 1986;

AP&L Energy Supply Procedure 214. "LRGO Preparation of Baseline

Technical and Quality Requirements for Commercial Grade Items in

l Q Systems," Revision 0, dated March 20, 1986;

• Administrative Procedure 1032.06, "Procurement Technical

Assistance," Revision 9 dated August 21, 1987;

• QC Operating Procedure QCO-5, "Purchase Requisition Review,

Receipt Inspection, sad Independent Material Testing," approved

May 22, 1987; and

QC Operating Procce a- .,0-9, "QC Discrepancy Identification and

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1sion 0, approved September 3, 1987.

The NRC inspector noted from this review that the governing site

procedure for handling significant discrepancies is Procedure 1000.08

Revision 23, and that Procedure QCO-4 refers to thia procedure for QC

identified significant discrepancies. Although A: u.:strative '

Procedure 1000.08 Revision 23, referenced 10 CFR le t 'l items in

the text, the NRC inspector did not note any clear , ance for

handling of these items or any specific reference to AP&L Energy

Supply Procedure 304 as the controlling document. In that this

particular review was performed subsequent to the exit interview and

insufficient time was available to discuss the matter with licensee

personnel, the adequacy of Administrative Proc.edure 1000.08,

Revision 23, with respect to control and reporting of potential

10 CFR Part 21 items is considered an unresolved item. (313/8804-05;

368/8804-06)

No additional violations or deviations were identified in this area

of inspection.

c. Specification of Applicability of 10 CFR Part 21 in Procurement

bocuments

The NRC inspector selected five purchase orders (P0s) (i.e.,

P0s 26884, 42607, 20602, 36970, and 16968) for safety-related

materials and components in order to determine whether the

requirements of 10 CFR Part 21 were being appropriately imposed on

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AP&L vendors. With the exception of PO 20602, the initial P0s i

l imposed 10 CFR Part 21 on the vendors. In the case of PO 20602 to 1

Hub, Inc. for ASME Section III Code, Class 3 flanges and fittings,

the initial PO dated July 7, 1987, did not specify the applicability

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of 10 CFR P. t 21. A supplement to the P0 was issued, however, on

July 10,196 , which identified that the applicability of 10 CFR

Part 21 had been omitted in error.

No violations or deviations were identified in this area of

inspection.

d. Review of 10 CFR Part 21 Evaluations

The NRC inspector reviewed two closed evaluations (Concern Log

Nos.1-28 and I-34) which had been determined to be not reportable by

the licensee. The NRC inspector concurred with the licensee's

evaluation results. A review was also performed of Concern Log

No. I-19, which is the item identified above as being received on

November 2, 1983, and in a currently open status. From the available

information, the NRC inspector did not identify a condition which

would appear to be reportable under 10 CFR Part 21.

No violations or deviations were identified in this area of.

inspection.

8. Unresolved Item

Unre50'ved items are matters about which more information is required in

order to ascertain whether or not the items are acceptable, violations,

or deviations. The following three unresolved items were discussed in

this report:

Paragraph Item Subject

3 368/8804-03 Adequacy of calculation

4 313/8804-02 Correction of Unit 1 PASS leakage path

7 313/8804-05; Adiquacy of procedure with respect to  :

368/8804-06 10 CFR Part 21 controls l

9. Exit Interview l

The NRC inspectors met with licensee representatives, denoted in

paragraph 1, on March 4 and March 10, 1988, and summarized the scope and

findings of the inspection.  ;

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