ML20195D760

Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,January-March 1988.(White Book)

ML20195D760
Person / Time
Issue date:	05/31/1988
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0040, NUREG-0040-V12-N01, NUREG-40, NUREG-40-V12-N1, NUDOCS 8806230188
Download: ML20195D760 (131)

Text

.

NUREG 0040 Vol.12, No.1 I

l LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT QUARTERLY REPORT JANUARY 1988 - MARCH 1988 l

UNlTED STATES NUCLEAR REGULATORY COMMISSION l

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Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 k

Washington, D.C. 20013-7082 A year's subscription consists of 4 issues for this publication.

Single copies of this publicatiori are available from National Technical Information Service, Springfield, VA 22161 l

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I NUREG 0040 Vol.12, No.1 LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT QUARTERLY REPORT JANUARY 1988 MARCH 1988 Date Pu ish d ay 198 Division of Reactor Ins..

and Safeguards Office of Nuclear Reactc, negulation U.S. Nuclear Regulatory Commission Washington, DC 20555 p>.....u,

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CONTENTS PAGE 1.

Preface y

2.

Reporting Format yjj 3.

Inspectors Reports 1

4 Selected Information Notices 91 5.

Index 123 6.

Table 6f Vendor Inspection Reports Related to Reactor Plants 125 iii

PREFACE A fundamental premise of the Nuclear Regulatory Commission's (NRC) nuclear facility licensing and inspection program is that licensees are responsible for the proper construction and safe operation of their nuclear power plants.

The total government-industry system for the inspection of nuclear facilities has been designed to provide for multipla levels of inspection and verification.

Licensees, contractors, and vendors each participate in a quality verification process in accordance with requirements prescribed by, or consistent with, NRC rules and regulations. The NRC inspects to determine whether its requirements are being met by a licensee and his contractors, while the great bulk of the inspection activity is performed by the industry within the framework of ongoing quality verification programs.

In implementing this multilayered approach, a licensee is responsible for developing a detailed quality assurance (QA) plan. This plan includes the QA programs of the licensee's contractors and vendors. The NRC reviews the licensee's and contractor's QA plans to determine that implementation of the proposed QA program would be satisfactory and responsive to NRC regulations.

In the case of the principal licensee contractors, such as nuclear steam supply system designers and architect engineering firins, the NRC encourages

_ submittal of a description of corporate-wide QA programs for review and acceptance by the NRC. Once accepted by NRC, a corporate QA program of a licensee's contractor will be acceptable for all license applications that incorporate the program by reference in a Safety Analysis Report (SAR).

In such cases, a contractors's QA program will not be reviewed by the NRC as part of the licensing review process, provided that the incorporation in the SAR is without change or modification.

However, new or revised regulations, Regulatory Guides, or Standard Review Plans affecting QA program controls may be applied by the NRC to prey wusly accepted QA programs.

When design and construction activities were high, firms designing nuclear steam supply systems, architect engineering firms designing nuclear power plants, and certain selected major equipment vendors were inspected on a regular basis by NRC to ase.ertain through direct observation of selected activities whether these design firms and vendors were satisfactorily implementing the accepted QA program.

However, with the substantial decline of new plant design activities, the inspection of QA program implementation has been deamphasized.

Instead, the NRC vendor inspection focus has been I

shifted to vendor activities associated with nuclear plant operatisn, maintenance, and nodifications.

Inspection emphasis in now placed on the quality of the vendor products including hardware fabrication, licensee-L v

vendor interfaces, environmental qualification of equipment, and equipment problems found during operation and corrective action.

If nonconformances t

with NRC requirements and regulations are found, the inspected organization is requested to take appropriate corrective action and to institute preventive measures to preclude recurrence.

If generic implications are identified, NRC assures that affected licensees are expeditiously informed.

In addition to the above, the Vendor Program Branch has begun inspections at l

licensee facilities covering the areas of procurement of replacement parts for use in safety-related systems and licensee / vendor interface programs as requested in Generic Letter 83-28.

This edition of the White Book contains copies of the inspection reports of inspections completed to date.

Subsequent issues will contain those reports that are issued in the quarterly report period covered by that White Book.

In the past, NRC issued confirming letters to the principal contractors to indicate that NRC inspections have confirmed satisfactory implementation of the accepted QA programs. Licensees and applicants could, at their option, use the letters to fulfill their obligation under 10 CFR 50 Appendix B, l

Criterion VII, that requires them to perform initial source evaluation audits l

and subsequent periodic audits to verify QA program implementation.

However, based on the above described change in nuclear plant design and construction activities, NRC will no longer issue confirming letters to principal contractors since future NRC vendor program inspections will focus on selected areas rather than addressing the implenientation of their respective QA programs. Therefore, l

confirming letters that have already exceeded their three year effective period will not be renewed.

Confirming letters issued less than three years ago will remain in effect until the stated effective period expires. Therefore, as the confirming letters expire, licensees and applicants will no longer be allowed to take credit for the NRC acceptance of the implementation of a principal contractor's QA program.

Licensees continue to be responsible for the conduct of initial source evaluation audits and subsequent periodic audits to verify QA program implementation.

The White Book will continue to be published and will contain copies of all vendor inspections issued during the calendar quarter specified. The vendor inspection reports list the nuclear facilities to which the results are applicable thereby informing licensees and vendors of potential problems.

In addition, the affected NRC Regional Offices are notified of any significant problem areas that may require special attention.

The White Book also con-tains copies of 1&E Information Notices, concerning vendor issues released during the calendar quarter.

The White Book contains information normally used to establish a "qualified suppliers" list; however, the information contained in this document is not adequate nor is it intended to stond by itself as a basis for qualification of suppliers.

Correspondence with contractors and vendors relative to the inspection data contained in the White Book is placed in the USNRC Public Document Room, located in Washington, D.C.

vi

ORGAllIZATION:

COMPANY, DIVISION CITY, STATE REPORT INSPECTION INSPECTION NO.: Docket / Year / Sequence DATE:

0N-SITE HOURS:

CORRESPONDENCE ADDRESS:

Corporate Name Division ATTN:

Name/ Title Address City, State Zip Code ORGANIZATIONAL CONTACT: Name/ Title TELEPHONE NUMBER:

Telephone Number NUCLEAR INDUSTRY ACTIVITY:

Description of type of components, equipment, or services supplied.

ASSIGNED INSPECTOR:

Name/ Vendor Program Branch Section Date OTHER INSPECTOR (S):

Name/ Vendor Program Branch Section APPROVED BY:

Name/ Chief - Section/ Vendor Program Branch Date INSPECTION BASES AND SCOPE:

A.

BASES:

Pertain to the inspection criteria that are applicable to the activity being inspected; i.e.,10 CFR Part 21, Appendix B to 10 CFR Part 50 and Safety Analysis Report or Topical Report comitments.

B.

SCOPE: Sumarizes the specific areas that were reviewed, and/or identi-fies plant systems, equipment or specific components that were inspected.

l For reactive (identified problem) inspections, the scope summarizes the problem that caused the inspection to be performed.

PLANT SITE APPLICABILITY:

List piant name and docket numbers of licensed facilities for which equipment, services, or records were examined during the inspection.

vii

ORGANIZATION: ORGANIZATION CITY, STATE REPORT INSPECTION N0.:

RESULTS:

PAGE 2 of 2 A.

VIOLATIONS:

Shown here are any inspection results determined to be in violation of Federal Regulations (such as 10 CFR Part 21) that are applicable to the organization being inspected.

B.

NONCONF0 PRANCES: Shown here are any inspection results determined to be in nonconformance with applicable comitments to NRC requirements.

In addition to identifying the applicable NRC requirements, the specific industry codes and standards, company QA manual sections, or operating procedures whicii are used to implement these commitments may be referenced.

C.

UNRESOLVED ITEMS:

Shown here are inspection results about which more l

informati'.a is required in order to determine whether they are acceptable l

items or whether a violation or nonconformance may exist. Such items will i

be resolved during subsequent inspections.

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D.

STATUS OF PREVIOUS INSPECTION FINDINGS: This section is used to identify the status of previously identified violations, items of nonconformance, and/or unresolved items until they are closed by appropriate action.

For all such items, and if closed, include a brief statement concerning action which closed the item.

If this section is omitted, all previous inspection findings have been closed.

E.

INSPECTION FINDINGS AND OTHER COMMENTS: This section is used to provide significant information concerning the inspection areas identified under "Inspection Scope." Included are such items as mitigating circumstances concerning a violation or nonconformance, or statements concerning the limitations or depth of inspection (sample size, type of review performed and special circumstances or concerns identified for possible followup).

For reactive inspections, this section will be used to summarize the disposition or status of the condition of event which caused the inspection to be performed.

F.

PERSONS CONTACTED: Typed, Name, Title

• present during exit meeting SAMPLE PAGE (EXPLANATION OF FORMAT AND TERMIN0 LOGY) viii

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INSPECTORS REPORTS i

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ORGANIZAT10N: ALLEGHENY LUDLUM CORPORATION CLAREMORE, OKLAH0MA REPORT INSPECTION INSPECTION N0.: 99902007/88-01 DATE:

01/4-6/88 ON-STTF MnitpS-9n CORRESPONDENCE ADDRESS: Allegheny Ludlum Corporation Tubular Products Division ATTN: William G. Bieber General Manager Post Office Box 948 Claremore, Oklahoma 74018-0948 ORGANIZATIONAL CONTACT: James Pepin, Technical Department Manager TELEPHONE NUMBER:

(918) 341-8711 NUCLEAR INDUSTRY ACTIVITY:

Fabricated E-Brite 26-1 alloy tubing for Engineers and Fabricators Company (EFC) and stainless 304L tubing for Yuba Heat Transfer Corporation both in 1981.

bd\\Adh

-88 ASSIGNED INSPECTOR:

4. V. Conway, Prograr Development and Reactive Date

\\1hspection Section DRIS)

OTHER INSPECTOR:

C. D. Sellers, M terials Engineering Branch APPROVED BY:

E~ ~

E. T. Baker, Acting Chief, PDRIS, Vendor Inspection Date Branch INSPECTION BASES AND SCOPE:

A.

BASES:

10 CFR Part, 50, Appendix B and 10 CFR Part 21.

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B.

SCOPE: The purpose of the inspection was to review the QA records and l

evaluate the process used to fabricate the E-Brite 26-1 tubing for the shutdwn cooling heat exchangers at Arkansas Nuclear One Unit 2 in 1981.

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1 PLANT SITE APPLICABILITY:

Shutdown cooling heat exchangers-Arkansas Nuclear One, Unit 2 (50-368).

1

ORGANIZATION:

ALLEGHENY LUDLUM CORPORATION CLAREMORE, OKLAHOMA REPORT INSPECTION NO : 99902007/88 RESULTS:

F AGE 2 of 5 A.

VIOLATIONS:

None.

B.

NONCONFORMANCES:

None.

C.

UNRESOLVED ITEMS:

None.

D.

STATUS OF PREVIOUS INSPECTION FINDINGS:

None.

This was the first inspection at this facility.

E.

OTHER FINDINGS OR COMMENTS:

Tubular Products Division (TPD) 1.

Allegheny Ludlum (AL) fabricates tubular products at their facility in Claremore, Oklahoma, which was built in 1978.

The original TPD, located in Wallingford, Connecticut, was phased out at the end of 1984 The TPD produces alloy steel tubing includfrg a number of ferritic grades designated E-Brite 26-1, AL 29-4C, and AL 29-4-2.

Based on a review of procurement and shipping records for tubing fabricated at both Wallingford and Claremore and af scussions with knowledgeable personnel, it has been determined that the only nuclear tubing fabricated by AL was in 1981.

This included E-Brite 26-1 alloy tubing for EFC and stainless 304L tubing for Yuba Heat Transfer.

Tube skelp from AL's Flat Rolled Products Division is' received by the TPD.

The skelp is roll formed into a circular shape un eight stand tube mills.

Continuous welding (TIG with argon-helium shielding) is done on 10 weld mills.

The weld bead is cold worked to produce a smooth surface on the 00 and ID.

In-line eddy current reverse bend) g and destructive tests (e.g., flatten, flare, andare performe (ET) monitorin weld.

Annealing with in-line induction heating / water quench or off-line bright annealing guarantees that both the parent metal and weld metal meet the mechanical and corrosion resistance requirements of the material specification.

Bright annealing is performed in a 3-zone furnace that has an atmosphere of either dissociated ammt.ifum 2

ORGANIZATION:

ALLEGHENY LUDLUM CORPORATION CLAREMORE, OKLAHOMA REPORT INSPECTION NO.: 99902007/88-01 RESULTS:

F AGE 3 of 5 for austenitic alloys or hydrogen for ferritic alloys.

This is followed by forced atmosphere cooling.

The gages and heat treat charts on the annealing furnace are calibrated by Honeywell.

ET of all tubing is done on three finishing lines. All tubing is continuously line stenciled per the material specification. Heat, bundle, and ET personnel numbers are included in the identification.

The Level II examiner also performs three mechanical tests (flatten, flange, and reverse bend).

Finishing also includes cutting, deburring and cleaning and may include polishing to remove surface oxides.

Stretch or rotary straightening assures straight tube lengths.

Pneumatic and' hydrostatic tests are performed on all tubes.

The gages used for these tests are sent to an cutside vendor for calibra-tion.

Final inspection includes visual plus an OD and ID check with calibrated micrometers and ring gages.

If the tubing undergoes U-bending, the ID is cleaned if lubricant has been used.

Following hending, the tubing is electric resistance stress annealed. A temp-late is used to check the bend radius.

The U-bent tubes are final cut and pc.ckaged for shipment.

The laboratory confinns the mechanical tests performed by the Level II examiner and performs tensile and hardness tests in acecrdance with the material specification. Corrosion tests are conducted if specified by the customer or the specification.

In addition to 100 percent ET, ultrasonic testing is also performed if it is required by the customer.

2.

E-Brite 26-1 The NRC inspector reviewed the QA records pertaining to the E-Brite 26-1 alloy tubing fabricated for EFC in 1981. EFC's Purchase Order (P0) 72257-D dated December 11, 1980 was for 166-3/4"0D x 16 BWG and 1332-3/4"0D x 18 BWG welded tubing to ASME SA-268 Grade TP XM 27.

The P0 contained supplementary material requirements (SMR) and required that procedures for ET, annealing, tube cleaning, tube bending, hydrotest, marking and packaging be approved by EFC prior to manufacturing.

The SMR also designated witness / hold points and specified that a mill test report include "certification that material conforms to 1980...ASME Code Section III, Class 2."

A TWX dated Jaruary 14, 1981 to the TPD indicated that EFC had approved nine procedures for the fabrication of the E-Brite tubes.

Seven TPD certified material test reports (CMTR) for the tubing were 3

ORGANIZATION: ALLEGHENY LUDLUM CORPORATION CLAREMORE, OKLAHOMA REPORT INSPECTION NO.: 99902007/88-01 RESULTS:

f AGE 4 of 5 signed by the QC Supervisor and referenced "1980 ASME Section III Class 2 and Section II SA268." During the inspection, tha TPD could not produce any documentation to sht that EFC had performed an audit /

1 survey to show that TPD had a QA program that met the requirements of NCA-3800 of Section III of the ASME Code.

Following the inspec-tion, TPD contacted EFC and received a copy of a December 18-19, 1980 survey conducted by EFC of the TPD in Claremore, Oklahoma.

The Nuclear Survey Report was later received and reviewed by the NRC inspector, and EFC stated that TPD had "an adequate Quality Systems Program and adequate means for its implementation."

A review of the Wallingford TPD's Quality Assurance Policy and Procedures Manual, dated December 1,1980 indicated it was patterned after the 18 criteria of Appendix B to 10 CFR Part 50.

This manual, along with the Quality System Program (QSP) were in effect when the E-Brite tubes for EFC wero fabricated.

The QSP summarized the QA program at the Claremore facility and addressed five major areas-Quality Organization; Control of Operations and Materials; Examina-tions, Test and Reports; Control of Measuring and Testing Equipment; and Audits and Corrective Action.

The NRC inspector reviewed a detailed processing plan for the E-Brite tubes which is completed prior to the issue of a mill ticket (shop traveler).

The plan is reviewed and approved by the QC Supervisor and Plant Manager.

The plan documents product information and specific activities in welding, heat treatment, straightline cutting, ET, pneumatic testing, visual inspection, U-bending, ID cleaning, stress relieving, and hydrostatic test.

A review of several shop travelers indicated that they included heat numbers and grade, required operations and tests, appropriate QA specifications /proce-dures, and mandatory hold and/or witness points.

The nine procedures referenced on the TPD CMTRs were reviewed.

The procedures included Heat Treating-Bright Annealing, Eddy Current Test Method, Supplemental Tube ID Cleaning, U-Tube Visual Examination Method, Process Control Dimensional Specifications (Straight and U-Bend Tubing), Hydrostatic Test, Stenciling Specification, Packaging, and Heat Treating-Stress Relieving Resistance Anneal.

3.

Yuba Heat Transfer Corporation (YHTC)

A review of procurement records and discussions with knowledgeable personnel at TPD indicates that the only nuclear work other than the E-Brite tubing for EFC was for YHTC.

YHTC P0 No. 81-3161 dated 4

ORGANIZATION: ALLEGHENY LUDLUM CORPORATION CLAREMORE, OKLAHOMA REPORT INSPECTION NO.:

99902007/88-01 RESULTS:

f AGE 5 of 5 July 17, 1981 was for 690-3/4"0D x 20 BWG 304L stainless steel tubes to material specification SA249.

The material was ordered to Section III, Class 3 of the 1981 edition of the ASME Code.

TPD's CMTR dated September 17, 1981 certified that the material met the requiren'ents of NCA-3800 and SA249-304L.

Other documents reviewed included a detailed processing plan, U-tube schedule, heat-treat charts, test report from Midstates Analytical Laboratories who performed tensile tests and metallographic examination, and nine tube mill test reports which showed that the flatten, flange, and reverse bend had beJn performed.

4.

.fg E The NRC inspector reviewed TPD's "Practice for Oualification and Certification of NDT Personnel" dated August 15, 1980.

The qualifica-tion records for seven examiners (2-Level III and 5-Level II) who were on the TPD payroll during the fabrication of the E-Brite tubes in 1981 were reviewed.

The records appeared to be consistent with the requirements of SNT-TC-1A.

F.

PERSONS CONTACTED

• William G. Bieber, General Manager

• James J. Pepin, Technical Department Manager James Hyslop, Marketing Manager Glenna Pennington, QA Technician

• Attended exit meeting.

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ORGANIZATION: AMERICAN INSULATED WlRE CORPORATION PAWTUCKET, RHODE ISLAND REPORT INSPECTION INSPECTION NO.: 99900399/88-01 DATES: 02/24-25/88 OM-SITE H0tlRS-PR CORRESPONDENCE ADDRESS: American Insulated Wire Corporation ATTN: Mr. I. S. Galkin President 36 Freeman Street Pawtucket, Rhode Island 02862 ORGANIZATIONAL CONTACT: Mr. A. Hadotan, Assistant Quality Assurance Manager TELEPHONE NUMBER:

(401) 726-0700 NUCLEAR INDUSTRY ACTIVITY:

None since 1984.

ASSIGNED INSPECTOR:

d/u d M [ M / N N i /

d N-ff R. N. M61st, Special Projects Inspection Section Date (SPIS)

OTHER INSPECTOR (S):

G. T. Hubbard, Office of Special Projects, TVA Projects APPROVED BY:

b NL(if' D31-N U. Potapovs, Chief, SPIS Vendor Inspection Branch Date INSPECTION BASES AND SCOPE:

A.

BASES:

10 CFR Part 50 Appandix 0 and 10 CFR Part 21.

B.

SCOPE: The purpose o' this inspection was to review and evaluate i

l American Insulated Wire sarporation (AIW) manufacturing, test and i

quality assurance (QA) records to verify that proper controls were l

implemented during all phases of the manufacture of silicone rubber (SR) insulated cables which are installed at Sequoyah and Watts Bar nuclear power olants.

PLANT SITE APPLICABILITY: Sequoyah 1/2 (50-327/50-328), Watts Bar 1/2 (50-390/

50-391),

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ORGANIZATION: AMERICAN 1NSULATED WIRE CORPORATION PAWTUCKET, RH0DE ISLAND REPORT INSPECTION NO.:

99900399/88-01 RESULTS:

PAGF 7 of R A.

VIOLATIONS:

None.

B.

NONCONFORMANCES:

None.

C.

UNRESOLVED ITEMS:

None.

D.

STATUS OF PREVIOUS INSPECTION FINDINGS:

1.

(0 pen) Nonconformance (84-01, Item B.1):

AIW QA Manual only covered manufacturing activities and did not esta-blish procedures for any of the 17 applicable criteria necessary to provide control over the equipment qualification progran.

No nuclear activities have been performed since 1984, therefore, imple-mentation could not be verified during this inspection.

2.

(0 pen) Nonconformance (84-01, Item B.2 a, b, c, d) a.

No aging data was available for EPDM wire tested by Franklin Institute Research Laboratories (FIRL) which was reported by FIRL to be aged by AIW 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> at 150 C and 600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> at 150 C.

b.

The data source on AIW 5585-EPR 60 percent retention of original elongation used for Arrhenius plots to determine 40-year life at 60 C and at 90 C was not available.

The data was reported to j

have been supplied by DuPont, j

c.

Long term agina data per Arrhenius techniques, which was required by TVA Purchase Order 79K5-825342, was not available or maintained.

d.

AIW personnel were unable to provide test plans to the NRC inspector describing AIW tests (thermal and irradiation aging j

followed by flames tests) in response to Ebasco Specification 211-73 requirements. Moreover, the NRC inspector verified that test plans were not available for any of the qualifications reviewed.

j 8

ORGANIZATION: AMERICAN INSULATED WIRE CORPORATION PAk' TUCKET, RH0DE ISLAND REPORT INSPECTION NO : 99900399/88-01 RESULTS:

Par # 'dR AIW's responses to these nonconformances had been previously reviewed by the Vendor Inspection Branch and accepted.

However, no nuclear activities have been performed since 1984, therefore, implementation of corrective action could not be verified during this inspection.

3.

(0 pen) Noncunformance (84-01, item B.3) sontrary to AIW response letter to NRC report No. 9990099/80-01, tnere was no evidence that inprocess QC inspectors had been reinstructed as committed. AIW's response to this nonconformance was reviewed by the Vendor Inspection Branch and accepted. However, no nuclear activities have been performed since 1984, therefore, implementation could not be verified during this inspection.

E.

OTHER FINDINGS AND._ COMMENTS:

1.

Background:

On September 4, 1987, the Tennessee Valley Authority (TVA) notified the Nuclear Regulatory Commission (NRC), pursuant to the requirements of 10 CFR Part 21, of several failures of SR insulated cables during insitu high potential testing.

The insulation failures occurred on three different SR insulated single conductor No.14 cables manufac-tured by AIW Corporation and installed at TVA's Sequoyah Nuclear Power Plant. Subsequent testing by TVA resulted in three additional insulation failures in SR insulated cables manufactured by AIW.

AIW manufactured the SR insulated cables during October 1979 through January 1980 under contract with TVA. AIW has not manufactured SR insulated cables for other licensee's.

Two different formulations for insulation compound were used during manufacture of SR insulated cable.

Insulation compound, General Electric (GE) SE-9104, recommended by GE was used first for manufac-ture of 250,000 feet of SR insulated cable.

However, qualification test reports for this cable were found to be unacceptable to demon-strate qualification when reviewed by TVA. AIW personnel were not sure why the SR cable qualification reports were unacceptable but thought that the concerns related to IEEE 383 Flame Testing require-ments. A!W indicated that this cable was never shipped or used.

Insulation compound GE SE 725 was next used for manufacture of SR insulated cable. AIW produced a sample (no record at AIW to l

9

GRGANIZATION: AMERICAN INSULATED WIRE CORPORATION PAWTUCKET, RHODE ISLAND REPORT INSPECTION NO.: 99900399/88-01 RESULTS-ParJ a af R indicate length) and submitted the sample to FIRL for qualification testing.

FIRL report F-C5119 dated May 1979 titled, "Qualification Tests of Electric Wires in a Simulated Loss-of-Coolant-Accident (LOCA) Environment" was deemed acceptable by TVA.

Subsequently, AIW produced 250,000 feet of SR insulated cable using the same manufac-turing methods that were used for qualification samples and shipped the cable to Sequoyah and Watts Bar Power Plants.

Subsequent to the insitu high potential test failures at TVA, AIW conducted spark testing of SR insulated cable at 3,200 VAC on three returned reels from TVA on September 2, 1987. There were no failures during this testing.

2.

Review of Certified Inspection Test Reports (CITRs):

Since AlW does not consider inprocess manufacturing records to be Quality Records and they are not retained, the inspectors were unable to evaluate the process used by AIW to manufacture the SR insulated cable using the E SE725 insulation compound.

However, r

the inspectors were able to determine that AIW purchased the SR insulation compound from GE already mixed. AIW then put the compound in the extruder and extruded it on the conductor. AIW production and quality personnel involved in the manufacture of SR cables were no longer at AIW; therefore the inspectors were unable to discuss the production of the cable with personnel.

Even though manufacturing records were not available, the inspectors were able to review the AIW CITRs for the subject SR cables.

The CITRs are considered Quality Records by A1W and are retained to demonstrate compliance with contract and/or specification require-ments.

The NRC inspectors reviewed seven CITRs relating to ship-ments of SR insulation cable using the GE SE725 insulation compound.

It was determined by the NRC inspection team that two lots were produced. One lot was manufactured during the last three months of 1979 and another manufactured during the first two months of 1980.

The CITR review included a review of:

a.

Dimensional and physical test data sheets b.

Final electrical test reports j

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ORGANIZATION: AMERICAN INSULATED WIRE CORPORATION PAWTUCKET, RHODE ISLAND REPORT INSPECTION NO.-

99900399/88-01 RESulTS-PArJ R nf G c.

Electrical tests on wire d.

Flame test report Test report showing tests witnessed by TVA e.

f.

Tests perfornied in accordance with TVA specification 3419 which included the following:

(1) Heat resistance tests (2)

Initial dielectric strength test (3) Dielectric strength test while immersed (4) Horizontal flame test (5)

Leakage resistance test (6)

Dielectric strength test after bounding (7) Moisture resisting test on covering It was determined by the NRC inspectors that the test data conformed with the acceptance criteria in TVA specification 3419.

There was no apparent difference in the test data for the two lots.

Additionally the inspectors determined that TVA Quality Personnel had performed source inspection on the cables at AIW prior to ship-ment to TVA. The inspectors reviewed these records and identified no deficiencies.

F.

PERSONNEL CONTACTED:

l Aram Madoian - QA Manager l

Normand Cagnon - QC Manager Bob Fortin - Lab Manager R&D l

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ORGANIZATTON:

CIRCLEVILLE METAL WORKS CIRCLEVILLE, OHIO l

REPORT INSPECTION INSPECTION N0.:

99901075/88-01 DATES:

01/11-14/80 l0N-SITE H0llRS-4R

'ORRESPONDENCE ADDRESS:

Circleville Metal Works C

ATTN: Mr. Michael Hooper, President 305 Logan Street Post Office Box 410 Circleville, Ohio 43113 ORGANIZATIONAL CONTACT: Mr. Joe Riddlebarger, Quality Control Supervisor TELEPHONE NUMBER:

(614)474-6016 NUCLEAR INDUSTRY ACTIVITY:

This spent fuel transportation cask is the first nuclear product manufactured by Circleville Metal Works.

ASSIGNED INSPECTOR:

M1/L k bl

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audia M. Abbate, Prograin Dev lopment and Dat-Reactive Inspection Section (PDRIS)

OTHER INSPECTOR (S):

Carl Czajkcwski, Consultant APPROVED BY:

f _P Edward T. Bakcr, Acting C'hief, PDRIS, Vendor Inspection ate Branch INSPECTION BASES AND SCOPd:

l A.

BASES:

10 CFR Part 71, Subpart H and 10 CFR Part 21, 1

B.

SCOPE:

This inspection was performed at the request of the Australian l

government and provided an independent verification of the fabrication of the LHRL-120 spent fuel transportation cask.

The areas reviewed during the inspection included welding, procurement, NDE, testing and inspection, measurirg and test equipment, ronconforming items, training anc audits.

The inspection consisted of observation of work ano a record enview.

PLANT SITE APPLICABILITY:

Not applicable, i

1

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ORGANIZATION:

CIRCLEVILLE METAL WORKS CTRCLEVILLE, OHIO REPORT INSPECTION NO.: 99901075/88-01 RESULTS:

PAGE 2 of 7 A.

VIOLATIONS:

There were no violations identified curing the inspection.

B.

NONCONFORMANCES:

1.

Contrary to Paragraph 71.115(b) of Subpart H to 10 CFR Part 71, Paragraph 18.1 of CMW-QAP-1, "Qublity Assurance Program," Revision C, dated December 4, 1986, Paragraph 3.1 of the CMW Quality Control Manual, Revision 1, dated March 3, 1986, and Section 2 of the Edlow International Quality Assurance Plan, Revision 4, dated September 9, 1987, the Certification of Chemical and Physical Properties for Purchase Order (PO) C7/2017-01/48 had missing chemical and mechanical test data, the material was not ordered to an ASTM /ASME specification and, although already installed in two basket assemblies, the material had not been approved by the CMW Quality Control Supervisor (QCS)

(88-01-01).

2.

Contrary to Paragraph 71.115(a) of Subpart H to 10 CFR Part 71, and Paragraph 3.1(a.1) of the CMW Quality Control Manual, Revision 1, dated March 3, 1986, P0 40938-99-8-02-03 for aluminum was purchased to specification QQ-A-200B, while the Certificate of Compliance (C of C) was not reviewed by CMW to that specification (88-01-02).

3.

Contrary to Paragraph 71.125 of Subpart H to 10 CFR Part 71, Sections 9.la and b of the CMW Quality Control Manual, Revision 1, dated March 3, 1986 and Paragraph 5.4a of CMW Specification #2, "Liquid Penetrant Examination," Revision 1, dated May 7, 1986, temperature gauges used in liquid penetrant tests and in weld rod ovens were not controlled in the CMW calibration system (80-01-03).

C.

UNRESOLVED ITEMS:

No unresolved items were identified during the inspection.

D.

STATUS OF PREVIOUS INSPECTION FINDINGS:

Not applicable since this was the first NRC inspection of the CMW QA program.

14

{

ORGAHlZATION:

CIRCLEVILLE METAL WORKS CIRCLEVILL E, OHIO REPORT INSPECTION N0.*

99901075/88-01 RESULTS:

PAGE 3 of 7 E.

OTHER FINDINGS AND C0KMENTS:

i 1.

Background Information This inspection was performed at the request of the Australian Depart-ment of Transport.

It was conducted to provide an independent verification of the fabrication of the LHRL-120 spent fuel transper-tation cask.

This transoortation cask will be used to transport spent fuel of U.S. Orig from the Australian Atomic Energy Commission (AAEC) Lucas Heights Research Laboratory in New South Wales, Australia to the United States.

Edlow International Company (EIC) is providing overall service for the project, including the cask dnd transportation.

Eggers Ridihalgh Partners, Incorporated (ERP) designed the cask and is responsible for the fabrication, QA program and acceptance testing.

CMW is responsible for the actual fabrication of the cask.

This is the first nuclear component manufactured at CMJ.

2.

Entrance and Exit _ Meetings An entrance meeting was conducted on January 11, 1968 at the CMW facility in Circleville, Ohio.

The purpose and scope of the inspec-tion were discussed during this meeting.

During the exit meeting, conducted on January 14, 1988, the inspection findings and observa-tion, were sucraarized.

3.

Fabrication Control During the inspection, the inspectors observed work being performed on the cask. The work area was rcped off from the rest of the facility and the cask and fabrication material were covered in plastic when work was not being perforned.

CMW had one welder perform most of the work on the cask and basket a ssen.bl i e s.

Thc qualifications of all the welders wno did perform work on the cask and baskets were reviewed.

This review consisted of exaninine the Welders Leg, which described what welders were qualified to what procedure. Also reviewed were the Weld Procedure Specifications (WPS), Procedure Qualification Records (PQR), and fabrication drawings for the cask and hasket asser,blies.

During this review several observations cor.cernir.g welding were nated by the inspectcrs.

These are as follows:

15

ORGAK!ZATION: CIRCLEVILLE METAL WORKS CIRCLEVIL LE, Oli10 REPORT INSPECTION

,_,p0.-

99901075/88-01 RESULTS:

PAGE 4 of 7 Upon review of the basket assembly drawing CMW 4035-2, it was a.

observed that a notation is shown for tack welding nuts to the basket.

CNW stated that the two baskets which had already been shipped to Australia actually had fillet welds in lieu of tack welds, contradicting the drawing requirements.

Since the baskets had already been shipped, the inspectors had no way of verifying the technique used. Tack welding of nuts is not aodressed by the ASME Code and is an undefined procedure due to the lack of control of weld heat input.

Cf',W stated that fillet welds would be used on the two remaining baskets and that this would be noted on the Fabrication Centrol Sheets and fabrica-tion drawings.

b.

Although required by drawing, there was no procedure or certified weloer available at CMW to weld P-6 stainless steel to P-1 carbon steel.

This process was to be performed later in the fabrication process and had not yet been used on the cask.

CMW connitted to writing a procedure and qualifying a welder to cover tnis item, During the inspection it was noted that temporary attachments c.

were welded to the cask to reduce ovality. There were no proce-dures available at CMW for the removal and inspection of these attachments. The CMW QSC modified the Fabrication Control Sheet to require a liquid penetrant test on all temporary attachment welds after attachment removal. This was confirmed by the inspectors during the inspection.

d.

The inprocess inspections and liquid penetrant tests were per formed by the QCS.

The liquid penetrant test procedure, liquid penetrant test reports and the qualifications of the QCS were reviewed. 'lhe QCS had certified himself as a Level III in Nondestructive Examination (NDF.),

it is not general practice l

to certify oneself; rather, a designated officer usualiy certifies NDE personnel. During the inspection the CMW president certified the QCS as a level 111 after reviewing the qualificetions. Additionally, the QCS's eye examination require-ment had expired prior to the inspection. The QCS is required to have an eye examination annually. During the inspection, the OCS passeo an eye examination thus fulfilling the SNT-TC-1A requirement, 1

t 16

ORGANIZATION: CIRCLEVILLE METAL WORKS CIRCLEVILLE, OHIO a

REPORT INSPECTION 0

NO.-

99991075/88 01 RESULTS:

pAGE 6 of 7 The inspectors also witnessed liquid penetrant testing. The QCS adhered to the procedure and in one instance, upon discovery of indications due to surface conditions, the QCS rejected the welds l

until the welds were ground smooth and retested Upon retesting,

}

the welds were found acceptable.

]

The temperature gauge used during the liquid penetrant test to verify that the temperature of the test material was between 60'F and 125 F was not controlled under the CMW calibration system.

During the inspection the temperature gauges were removed from the calibration system ar.d the QC Manual was revised. CMW will rely on the manufacturer's guarantee of a 2 F accuracy since the gauges are used for qdalitative rather than quantitative measurements.

F Monconformance 88-01-03 was identified in this area.

CMW installed three portable weld rod ovens in the roped off area.

The weld rod contained in the ovens was to be used only for this project. This was done to segregate th: stainless steel weld rod used on this job from the other carbon and stainless steel weld rod used throughout the shop. Upon inspection of the ovens, it was noted thet the ovens were at three cifferent temperatures vacying from 175 F to 250'F and that the temperature gauges being used ir.

the ovens were not included in the CMW calibration systen (see nonconformance 88-01-03).

These temperature gauges were the same model as that used during the liquid penetrant test.

During the inspection, the weld rod manufacturer was contacted by CMW to obtain

^;

the correct weld rod storage tenperatures.

CMW was informed by the manufacturer that the weld rod should be stored in a warm dry place.

CHW feels that the 2 F accuracy range guaranteed by the gauge manufacturer will be sufficient for inuiCating oven temperature and therefore, removea the temperature gauges f rom the calibration system.

The inspectors observed hold tags on son.e material in ne shop.

These tags, although not nurbered, were traceable back to the nonconformity report (NCR) and the nold tag form. The inspectors did not identify any instance where treterial had on NCR writter on it and did not have a hold tag at'. ached.

ERP had imposed a requirement on CVW sta+irg that the chloride content of anything in contact with the cask would be less than 100 ppm.

In response to this requirement, CMW hao the Circleville water

i su,oply, marking pens used on cask material, and other materici in

tp 17 L

t

-0RGANIZATION: CIRCLEVILLE METAL K0PKS CTRCLEVILLE, CHIO REPORT INSPECTION NO.-

99901075/88-01 RESULTS:

PAGE G of 7 _

contact with the cask analyzed for chloride content.

The cert.jfica-tions for these tests were reviewed and found to be, for the most part, in compliance with 100 ppa limit.

h was noted by the inspectors, however, that in some cases when the standard deviation was added to the measured chloride content, the 100 ppni specifica-tion was exceeded, in aadition to the inspection of each of the fabrication steps on the Fabrication Control Sheet by the QCS, ERP has a list of audits of procedures to be performed during hold and witness points. The procedures to be audited and witnessed by ERP include material inspection, liquid penetrant tests and visual examinations cf welds, basket assembly, lead pour, y mma scan, and various leak tests, lift tests and acceptance tests.

Each of the audits is performed according to a che.klist the ERP QA manager fills out during the audit.

4.

Procurement The inspectors reviewed the procurement system at CMW.

Purchase Orders, C of Cs, test reports and receipt inspectiun forms were revieveu. During this review two nonconformances, 88-01-01 and 88-01-02, were identified in the area of purchased material control.

During the review cf the Certification of Chemical and Physical Properties fo.' PO 87/2017-01/48, it was noted that the material was apparently not purchased to an ASTM /ASME specification. The EIC Quality Assurance Plan states that the ASME BPV, Section 11 be used for material specif 4.ation reference. Additionally, some data for chemical tests and all the data for mechanical tests were missing. The certification had the QCS stamp on it, but the QCS had not approved or datad the certification. The material had been purchased by ERP and was used in No basket assemblies which have already been shipoed to Aus' ra',

During the inspection CMW obtained the missing chemSal d6.

and committed co obtaining samples from the lot and performing tensile, elongatior., yield strength and hardness tests. The results from the chemical and mechanical tests were reviewed after the inspection and the inspectors verified that the material met the chemical and physical soecifications for ACTM B241 material.

Additionally, during the review of a certified mill test report (CMTR) for 6061-T651 aluminum, it was noted that the material was purchased to soecification QQ-A-200B, but the CMTR was not reviewed i

to that specitication by the QCS.

CMW personnel obtained a copy of

.., ~

18

ORGAN 12ATION: CIRCLEVILLE METAL FORKS CIRCLEVILLE, OHIO REPORT INSPECTION NO.: 99g01075/88-01 RESULTS:

PAGE 7 of 7 QQ-A-200E (the latest revision) and a copy of ASME 8B-211 (a compara-ble specification) during the inspection and compared them to the CMTR.

The material was found to be acceptable.

5.

10 CFR Part_21 The inspectors reviewed CMW procedure, "Section 206 Nonconformity Report," Revision 1, dated November 3, 1987. This procedure outlines the steps to be taken when a nonconformity is identified as described in 10 CFR Part 21.

The inspectors noted that the procedure contained provisions for informing the licensee or purchaser (in this case ERP) of the nonconformance, however, it was unclear as to who notified the NRC. The inspo.ctors reviewed ERP's procedure, QAP 15.1, "Reporting of Defects and Nonconformance," Revision 0, dated June 1, 1986.

ERP's procedure did provide for notification of the NRC. During the inspec-tion, CMW issued revision 2 to the procedure which identified the CMW president as being responsible for verifying hRC notification.

6.

Audits For this job, no specific internal audits of the CMW QA progrem had been perforr.ed by CMW.

ERP, however, had reviewed and approved the CMW QA program prior to fabrication.

Subsequent to the inspection CMW committed to perform an internal audit before the LHRL-120 project was complettd and annually thareafter if more nuclear products are manufactured.

Only one externel audit of a supplier was performed for this job.

The QCS had obtained Qn program manuals for most of the remaining suppliers used to supply material for this job. Of those, most had suppliec' quality materials to CHW prior to this job with no history of supplying pccr quality materials.

Subsquent to the inspection CMW connitted to initiate a supplier survay log for potential suppliers and perform on-site audits when needed.

F.

PERSONS CONTACTED:

• Herman Crawford, QA lianager, Eggers Ridihalgh Partners, Incorporated

• Michdel Hooper, President, Circleville Metal Works (CMW)

• Joe Piddleherger, QSC, ChW

• Jerold Schaeper, Project Coordinator, CMW Jce Shomaktr, helder, CMW

• Howard Sct.el, QA Manager, Edlow Internationcl Compeny

• Attended ait meeting, i

19

ORGANIZATION: GENERAL ELECTRIC COMPANY WILMINGTON, NORTH CAROLINA l

i REPORT INSPECTION INSPECTION N0.:

99900003/87-01 DATES:

12/07-11/87 ON-SITE HOURS:

31 i

CORRESPONDENCE ADDRESS:

General Electric Company Nuclear Fuel and Components Manufacturing Facility ATTN: Mr. E. A. Lees, Manager Post Office Box 780 Wilmington, North Carolina 28402 ORGANIZATIONAL CONTACT: Mr. J. W. Currier, Jr.

TELEPHONE NUMBER:

(919) 343-5874 NUCLEAR INDUSTRY ACTIVITY; Nuclear fuel assembly and BWR core hardware supplier for GE designed reactors and fuel pellet fabrication for Babcock and Wilcox (B&W) fuel nssemblies.

ASSIGNED INSPECTOR:

e!sW 5 fg R. L. Ciliinb' erg, Program))tivelopment and Reactive Date Inspection Section (PCRiS)

OTHERINSPECTOR(S):

B. J. Mincher, Consultant APPROVED BY:

G C'V 8

E. T. Baker, Acting Chief, PORIS, Vendor Inspection Branch Date INSPECTION BASES AND SCOPE:

A.

BASES:

10 CFA 50, Appendix B,10 CFR Part 21.

B.

SCCPE:

Follow-up allegations related to activities in the Chemet ELo7atory ana Quality Assurance (QA) program implementation related to fabrication of GE fuel.

PLANT SITE APPLICABILITY: BWR facilities with fuel spplied by GE and B M facilities with fuel pellets supplied by GE.

21

4 ORGANIZATION: GENERAL ELECTRfC COMPANY WtLMINGTON, NORTH CAROLINA REPORT INSPECTION No.-

99900003/87-01 RESULTS:

PAGE 2 of 6 A.

VIOLATIONS:

None.

8.

NONCONFORMANCES_:

1.

Contrary to Criterion V of Appendix B to 10 CFR 50 and Section 1 of Char.geroom PROD No. 86.06, "General Lab Controlled Area Rules,"

Revision 0, dated January 20, 1987, a GE non-laboratory employee was chewing gum in the Chemet Laboratory.

(87-01-01) 2.

Contrary to Criterion V of Appendix B to 10 CFR 50 and Section 4.0 of the Foreword to Chemical Metallurgical 6 Spectrochemical (CMJS)

Manual, Revision 7, dated June 18, 1986, a pen and ink change was made to Analytical Test Method (ATM) 1.2.8.3, "Determination of Sintered Pellets and 00 Hydrogen in Zirconium Its Alloys, U0p 7

Gd,0 by Inert Gas Fusion," Revision 7 dated September 19, 1986, 2

with0ut initiating a formal revision to ATM 1.2.8.3, Revision 7.

(87-01-02) 3.

Contrary to Criterion V of Appendix 8 to 10 CFR 50, ATM 1.2.8.3, Revision 7 did not documenc the ste;- being used by the technician to clean the LECO Impulse Furnace (LIF).

(87-01-03) 4.

Contrary to Criterion V of Appendix B to 10 CFR 50, ATM 1.2.8.3, Revisior, 7, did not document the technician physically bumping the LIF to evenly distribute the flux material.

(87-01-04) 5.

Contrary to Criterion I of Appendix B to 10 CFR 50, ATM 1.2.8.3, Revision 7, did not provide guidance for the action taken by a technician when a peliet does not drop into the crucible in the LIF.

(87-01-05)

C.

UNRESOLVED ITEMS:

l None.

D.

STATUS OF PREVIOUS INSPECTION FINDINGS Not applicable.

l l

l 1

I 22 1

ORGANIZATION: GENERAL ELECTRIC COMPANY WILMINGTON, NORTH CAROLINA REPORT INSPECTION NO.: 99900003/87-01 RESULTS,:

PAGE S of 6 E.

INSPECTION FINDINGS AND OTHER COMMENTS:

1.

Entrance and Exit Meetings The NRC staff informed GE management representatives of the scope of the inspection during the entrance meeting on December 7,1987, and summar! zed the inspection findings and observations during the exit meeting on December 11, 1987, 2.

Allegation An allegation was received by the NRC concerning the GE Chemet Laboratory which stated tht lab technicians and supervisors eat food and chew gum in the laboratory and cheat on various analytical tests being performed in the laboratory.

Chewing gum in the labora-tory was substantiated as discussed in Section 3(d) below. The allegation also stated that false numbers were being put into a report by a technician instead of running the tests for hydrogen in U0, fuel pellets. The allegation also stated that technicians were chhating because they were completing their tests in a short period of time. The NRC inspectors could not substantiate these allegations because technicians testing for hydrogen in UO fuel pellets were observed to be adhering to procedural time req 61rements and to be enteiing correct numbers into a computer system instead of a report.

The computer system is identified as the LMCS system and is discussed in Section E.3.c. below.

3.

Chemet Laboratory The NRC inspectors reviewed test procedures and results and observed analytical work being performed in the Chemet Laboratory, a.

Procedure Review The following test procedures were reviewed: ATM 1.2.8.3, l

Revision 7; ATM 1.2.7.2, "Analysis of Backfill Gas," Revision 3, dated February 20,1985; SCP204, "Analysis of Backfill Gas at Wilmington," Revision 7, dated June 4, 1986; SCP205, "Rod MICS Controls," Revision 3, dated November 26, 1985; SCP215, I,

"Thermal Simulation," Revision 5, dated February 27, 1986; SCP226, "Thermal Simulation (Manual Meas.)," Revision 0, dated September 23, 1985; CMS 1.2.4.8, "Geometric Density Measurement of B&W Fuel Pellets for use with the Thermal Simulation Test,"

I 23

ORGANIZATION: GENERAL ELECTRIC COMPANY WILMINGTON, NORTH CAROLINA REFORT INSPECTION N0.: 9990C003/87-01 RESULTS:

PAGE 4 of 6 Revision 3, dated September 11, 1986; and the LEC0 Instruction Manual for the RH-1EN Hydrogen Determinator Model 772-800/BCD, Copyright LECO Corporation, 1975.

The procedural review revealed that a pen and ink change had been made on ATM 1.E.8.3, Revision 7, which deleted the word "glass" from the procedure. A revised procedure had not been formally issued a; required by the CMS Manual, Revision 7.

(See nonconfonnance 87-01-02.)

b.

Observation of Analyses The NRC inspectors observed technicians performing analyses utilizing the procedures discussed above.

(1) Technicians were observed cleaning the LIF between analyses of hydrogen in U0 fuel pellets. AT11 1.2.8.3., Revision 7, did not provi<!e a6y instructions for this cleaning step.

(See nonconformance 87-01-03.)

(2) Technicians were observed bumping the LIF with their hands during one step in the hydrogen analysis. The technicians stated that bumping the LIF distributes the flux material.

ATM 1.2.8.3, Revision 7, did not provide any instructions for burnping the furnace.

(See nonconformance 87-01-04.)

(3) After the completien of a hydrogen analysis, the technician discovered that the U02 pellet had not dropped into the crucible in the LIF. The technician explained that the pellet does not drop into the crucible during approximately 10 percent of the analyses. When the pellet does not drop, the technician reruns the analysis but ATM 1.2.8.3, instructions to the Revision 7 did not provide any(See nonconfonriance 87-01-05.)

technician on stuck pellets.

l (4) ATM 1.2.8.3, Revision 7 requires that verification samples be analyzed after 25 UO,3 pellets have been analyzed for hydrogen. The inspectobs detennined that two technicians 1

analyzed verification samples in accordance with the procedure and the LMCS data system acceptance criteria.

(See Section E.3.c. below.)

1 (5) The inspectors determined that technicians were performing l

analyses in accoroance with CMS 1.2.7.2, Revision 3 and I

CMS 1.2.4.8, Revision 3.

l 24

ORGANIZA110N: GENERAL ELECTRIC COMPANY WILMINGTON, NORTH CAROLINA REPORT INSPECTION NO.: 99900003/87-01 RESULTS:

PAGE 5 of 6 4

c.

Computerizei Data and Calibration / Verification System The Chemet Laboratory uses a system identified as LMCS which employs computer tracking of data and calibration requirements.

The inspectors reviewed the calibration histories for equipment and the verification sample data used for 2 hydrogen analyzers.

The LMCS system requires that the hydrogen analyzers be calibra-ted prior to analyzing U0 fuel pellets for hydrogen content.

2 The calibration data must be entered into the LMCS system or the LMCS system will not accept hydrogen test results. A technician could enter false hydrogen test results only after false calibration data was entered. The NRC inspectors cbserved only correct calibration data and hydrogen test results being entered into the LMCS system by two technicians.

d.

Radiation Protection The NRC inspectors observed a GE non-laboratory employee chewing gum in the Chemet Laboratory.

The GE employee discontinued chewing 9.m when it was brought to his attention, and he stated that his action deviated from Changeroom PROD No. 86.06, Revision 0.

This substantiates part of the allegation discussed in Section 2. above.

(See nonconformance 87-01-01.)

e.

Conclusions The technicians knew their analytical responsibilities as evidenced by their ability to achieve hydrogen results with techniques that were not documented in ATM 1.2.8.3, Revision 7.

ATM 1.2.8.0, "Determination of Hydrogen in Zirconium, its Alloys, U0 Sintered Pellets, and U07 - Gd 0g Pellets by inert Gas 7

9 Fusion," Revision 6, dated OGcember 10, 1967, tras prepared during the inspection to iespond to nonconformances 87-01-02 through 87-01-05. The revised procedure is responsive to the nonconfor-mances, but had not been approved prior to the exit meeting en December 11, 1987.

4.

Fuel Manufacturing The NRC inspectors observed various steps in fuel pellet grinding and inspection, end cap weld inspection, and fuel bundle packaging to assure that processes were being performed according to written procedures by qualified operators.

Fuel manufacturing steps were perfonneJ in accordance with the following procedures: QCOR 3.1.4.1, 25

ORGANIZATION: GENERAL ELECTRlC COMPANY WILMINGTON, NORTH CAROLINA REPORT INSPECTION N0.: 99900003/87-01 RESULTS:

PAGE 6 of 6 "Pellet Grinding," Revision 11, dated November 1,1985; QC 113.2.3.1, "Pellet Grinding," Revision 21, dated December 1,1986; QNFD 1292, "Pellet Diameter Overcheck," Revision 1, dated October 31, 1987; QCOR 7.2.7.2, "First Weld Inspection," Revision 19, dated October 28, 1987; and QCOR 5.1.10, "Packaging of Fuel Bundles,"

Revision 16, dated July 1, 1986.

No items of nonconformance or unresolved items were identified.

5.

Internal Audits The inspectors reviewed internal audit reports for 1987 to determine confonnance with the GE QA program.

Internal audits are performed in accordance with P/P 30-13 "Internal Quality Audit Program,"

Revision 9, dated October 29, 1985.

Production and inspection areas are audited at an interval of at least once each calendar year.

Personnel performing internal audits must neet the requirements of QASAR 320-150.01 "Training and Qualification of QA Internal Audit Personnel," Revision 7, dated July 14, 1986. The inspectors reviewed the qualification and training files for auditors performing the 1987 internal audits and determined that the auditors met the requirements of QASAR 320-150.01, Revision 7.

F.

PERSONS CONTACTED:

D. Carroll

• J. Currier P. Custer H. Fields R. Hoffm6n

• P. Jasinski D. Joyner

• J. Liberman

• W.

McMahon S. Murray C. Ogle R. Patterson

• D. Pensinger B. Pierce J. Pierce l

• E. Schaefer

• L. Sheely G. Smithwick J. Wolf

• Attended exit meeting.

26 l

l

ORGANIZATION: GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA REPORT It!SPECTION INSPECTIUN N0.: 99900403/87-06 DATES:

11/09-12/87 ON-SITE HOURS:

117 CORRESPONDENCE ADDRESS:

GE Nuclear Energy ATTN:

Dr. Bertran Wolfe, Vice President and General Manager 175 Curtner Avenue San Juse, California 95125 ORGANIZATIONAL CONTACT: Mr. J. J. Fox, Senior Program Manager TELEPHONE fiUMBER:

(408) 925-6195 NUCLEAR INDUSTRY ACTIVITY:

GE huclear Engery is engaged in furnishing engineering services for domestic and foreign nuclear power plants.

O A

i ASSIGNED INSPECTOR:

b. k.p h b l2 hk R. L. Pettis, Special Projects Inspection Section Date OTHER INSPECTOR (S):

R. P. McIntyre, SPIS W. P. Haass, SPIS P. E

, Consultant b

APPROVED BY:

Mlh SIan.A 2-2.C- @

M U. Potapovs. Chief,~ SFIS, Vendor Inspection Branch Date INSPECTION BASES AND SCOPE:

A.

BASES:

10 CFR Part 21 and 10 CFR Part 50.

B.

SCOPE: The purpose of this follow-up inspection was to review allega-tions involving potential deficiencies in design control activities within the Quality Assurance program at Ge San Jose, during the period March 1978 to April 1982.

In addition, the status of previous inspection findings was also reviewed.

PLANT SITE APPLICABILITY:

Potentially multiple plant sites, including River Bend, TVA Units 17-22 (identified by GE as cancelled), Perry 1/2, Nine Mile Poirt 2, Hope Creek 1/2, Grand Gulf 1/2, Limerick, Clinton, and Susquehanna 1/2.

27

ORGANIZATION: GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA REPORT INSPECTION NO.-

99900403/07-06 RESULTS:

PAGE 2 of 30 A.

VIOLATIONS:

None.

B.

NONCONFORMANCES:

None.

C.

UNRESOLVED ITEMS:

None.

D.

STATUS OF PREVIOUS INSPECTION FINDINGS:

1.

(Closed) Nonconformance (87-03-01)

Documentation was unavailable during the inspection to demonstrate GE's extending the original seismic qualificaticn of the reactor mode switch, performed in 1978, tc a 1980 revised design configura-tion.

GE's response to this item of nonconformance identified that the mooe switch design changes were made on ECHs NJ21792 and NJ21793 dated December 15, 1980.

GE also stated that Loth the responsib*ie design engineer and the independent design verifier were aware of the recommendations made earlier by the GE problem review board that a new seism'.c analysis be performed to auclify the switch to the new design changes.

They also cencluded, after an analysis that was not formally documented, that the original design verification is not affected and so noted this statement on the above ECNs. A 'temo of Record" generated during the NRC Inspection on August 4, 1987 formnlly documented this analysis unc again concluded that the mode switch design changes did not affect the original seismic quclification.

This item is closed.

j 1.

(Closed) Unrescived Item (86-01-07)

GE Ergineering Practices and Procedure (EP&P) 5.38 Addendum 4, dated l

I December 1975, required that a tracking system and status log of deferred verifications be maintained.

The inspectors verified during the NRC 86-01 inspection that the first entry was made in the status 109 for deferred verifications in May 1977.

At that time, it 28

ORGANIZATION: GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA REPORT INSfECTION t!C. : 99900403/87-06 REStil.TS:

PAGE 3 of 30 could not be determined whether verifications had been deferred before May 1977 since the status log did not contain any deferred ferificctions entries prior to t)1at date.

GE provided additional documentation (several ERMs deferring verifica-tion from three saparate work units within GE) that indicated deferred verification activities were initiated as early as November,1975.

Although ducumentation did not exist to support the inclusion of these documents into the status log, GE stated they were incorporated but subsequently transferred to the Work Planning and Scheduling System (WPSS) for scheduling and tracking to completion.

The only documentation produced by GE to verify completion status of these docunents was a computer run from the Engineering Information System (EIS), dated April 4,1987, which indicated that 272 documents presently exist in the system as "U" (unverified). These documents are associated with ongoing design work in process today at GE San Jose.

GE's position was that since the drawings referenced on the ERMs were absent from this list, they must have been verified.

It was also pointed out by GE that the WPSS scheduling records prior to 1980, which would have demonstrated the documents tracking status and eventual closure, have been eliminated from the system data base. As a result, documentation did not exist to support the overall tracking status of these documents including the clearing of the referenced deferred verification.

In an attempt to demonstrate tracking of deferred verifications, GE committed to perform an extensive review of deferred verifications from inceptior through May 1977 to positively demonstrate closure of deferred design verifications. During an NRC review of this effort during the June 15-18, 1987 inspection, GE had reviewed all 15,300 Engineering Review Memorandums (ERMs) generated from inception to May 1977 to identify ERMs containing a deferred verification statement. As a result, 974 ERMs were identified which affected 3434 design documents.

A comnuter search of these documents, performed on GE's Engineering Information System (EIS), was then used to identify the current deferred verification status of the affected documents. A DBase III computer program was used by GE to produce a list of deferred verifications based on criteria established by the NRC during a previous NRC inspection.

This criteria was based on safety-related shippable components produced by GE NEB 0, San Jose, for use on domestic nuclear power plants.

This search produced approximately 150 design documents of which the NRC inspector selected six for further review by GE.

29

ORGANIZATION: GE NUCLEAR ENERGY SAN JOSE, CALIFORfilA REPORT INSPECT'ON

_ NO.-

99900403/87-06 RESULTS:

PAGE 4 of 30 Inspection Finding - GE's review of the six flRC selected documents consisted of a manual search of docurrentation necessary to demonstrate positive opening and closing of each deferred verification throughout the history of each document. The results of this review indicated that all deferred verifications made on these selected design documents were found to be completed and closed as required.

The inspectors reviewed in detail all the applicable documentation which a

demonstrated the opening and closing of deferred verifications I

throughout the documents history. This review included all revision levels of th design document and all ERlis or ECNs initiated per revision.

Ti e inspector also verified that documents were listed as

}

complete (C) on GEs Engineering Informatior. System (EIS).

For the 6 selected design documents, all design verifications deferred during 1974-1976 were completed prior to December 1976.

Eased on this review, this unresolved item is closed. Additional information concerning the review of the deferred verificction system and GE's actions are presented in Section E.2 of this report.

(_0 pen) Stokes Report Section 1.6 3.

Engineering Review Memorandums (ERMs)

"In the first week of November 1978, the followinJ line was part of an entry: Bill Millaro said either he would sign the ERMs or I (Sam) could forge his signature to them."

(Clarification added by Mr. Stokes.)

This iten was not addressed during this inspection.

4.

f0 pen)StokesReportSection1.7 Elementary Diogram Draf tine Effort "Continuing with a problem of similar nature on November 14, 1970, e letter to C. W. Part on the subject of the CliV connectinn has an interestir,g paragraph.

It seems that the CNV elementary diagran drafting effort was sut, contracted to an outside firm, the Fcwer Division of C. F. Braun & Company, in Alhambra, Cali brnia. When completed, the diagrams were provided to the General Electric System Engineers for signature.

The syrtem Engineets felt that they were not being given sufficient tirie for review and refused to sign the documents.

The documents were later signed by the CSEE CNV Engineer, wi thout review."

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ORGANIZATION: GE liUCLGR EtiERGY SAfi JOSE, CALIFORNIA REPORT INSPECTION N0.:

9990040s/87-06 RESULTS:

PAGE 5 of 20 This item was not addressed during this inspection.

5.

(0 pen) Stokes Report Section 6.2 Un3uthorized Signature Changes "Nr. Milam wrote a letter to W. M. Barrentine on April 14, 1982 about unauthorized post signature changes.

In this letter, Mr. Milam states that R. L. Reghitto made an authorized change to ERM AML-2997 without Mr. Milam's knowledge and in direct ccnflict with specific instructions."

This item w6s not addressed during this inspection.

6.

(0 pen) Stokes Report Section 6.3 Letter to Management "On May 22, 1902, Mr. Milam wrote Mr. Barrentine a letter and included a copy of his work record while working for Mr. C. L. Cobler.

In this letter, Mr. Milam requested Mr. Barrentine to read about the on-going underworld of C&ID and says he tried to communicate some of these things to Mr. Barrentine en several occasions but was discouraged by Mr. Barrentine's managers and attitude. Mr. Milam says:

Since you no longer hold my form 38 (a standard threat).

I have nothing further to fear from either you or your conspiratoral managers.

I hope, by sending you this Record, to give you a glimpse into that hidden world of uncontrolled bootleg activity we all know so well.

Mr. Stokes also stated that Mr. Barrentine was the manager of the Nuclear Control & Instrumentation Product Design Operation (NC&lD) of (C&ID). He was Mr. Hart's, Mr. Cobler's, Mr. Reghitto's, Mr. Strambach's, Mr. Koslow's, and Mr. Wortham's supervisor.

l Mr. Milam had been notified of his layoff when this last letter was written and his reference to form 38 had to do with the constant l

threat of layoff if you did not go along with the system. He did not "

l l

This item was not addressed during this inspection.

31

ORGANIZATION: GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA REPORT INSPECTION N0.: 99900403/87-06 RESULTS:

PAGE 6 of 30 7.

(0 pen) Stokes Report Section 5.13 River Bend Excluded Equipment List "Mr. Milam's work record included a nonapproved form titled PWA No.

1229LD, Revision IJ for River Bend.

This document, which is dated February 5, 1982, was caused by an excluded equipment list which was sent to the utility, Gulf States Utilities Company, by the NRC.

The second page of this document states that there is no controlled tracking systen for vendor identification of these devices and that a complete item by item search of the entire River Bend database would be necessary. GE felt that the scope of such a search was prohibitive and furtherrore was not considered to be necessary.

Excluded equipment as referred to in this list is equipment which has been found at other facilities to be so deficient that plant safety is seriously in question. GE neither admitted nor denied that this equipment was installed at River Bend."

This item was not addressed during this inspection.

E.

OTHER FihDINGS AND OBSERVATIONS:

Background Information As stated previously, NRC Inspection Report Nos. 99000403/G6-01, 87-01, dnd 87-03 did not address all of the allegations raised by Mr. Milam and Mr. Stokes, but rather, e representative sample of potentially more significant allegations was selected for review.

oo-ver, all allega-tions received by the NRC are being addressed and will be documented in future inspection reports.

Previously, the area of deferred design verification was acdressed which represented Mr. Milom's major concerns (as noted during an NRC interview with Mr. Milam ir April 1986).

As stated in Section D.2 cf this repcrt, this item is closed.

1.

Kaowool vs. Sand Another crea potentially affecting the protection of control roon instiumentation in the event of a fire was covered in an i ~ ternal GE memorandum dated key 23, 1980 with regard to the fire stop design requirements for Grand Culf 1/2 and Clinton 1.

The nemorandun stated thet a combinatien of metel barriers and Kaownol, both covered with RTV Rubber would cerstitute the fire break design in the control roem under-the-floor cable troughs.

This memorandum also indicated concern about the inability of Kaowool to fill the 1

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CPGANIZATICN: GE NUCLEAR ENERGY SAN JOSE, CAllFORillA REPORT INSPECTICH NO.:

99000403/87-06 RESULTS:

DArp 7 M v cable interstices and that Kaowool may be tuo easily removed.

The memorandum also stated that unless specific HRC approval is obtained, this design approach may be unsatisfactory.

During early discussions between GE :nd NRC concerning fire puiection for under-the-floor cable troughs in the PGCC, one oesign concept that was considered involved filling the cable trough with sand.

After considerable discussion of the idea in 1976-1978, all parties agreed that the potential disadvantage outweighed the benefits of using sand as a fire stop, especially considering the low risk of fire occurrence and associated damage to control room cables.

Therefore, the concept cf filling the cable troughs with sand was not adopted by GE and was never required by NRC.

GE's current design of the fire stops is based upon test data obtained by GE during tests conducted at the University of Lalifornia at Berkley. The design concepts are presented in design concept document NED0-104C0A titled, "Power Generation Control Complex Design Criteria and Safety Evaluation." This document and amendments are referenced in the FSAk's of the GE plants utilizing PGCC equipment. The criteria for the fire stop material is referenced as 3 inch minimum of a refractory material.

The tests at the University of California are included as a reference in this occument.

The refractory material used in these tests was No. 20 sand. A refractory blanket material is currently utilized as a fire stop material in openings which do not have cables passing through the opening. However, in the areas where cables are present, an RTV foam material is applied as the fire stop material and sealant between the cable trough and is utilized throughout current plants as a fire stop.

GE stated the existence of an analysis documenting the acceptability of the RTV foam material in lieu of using No. 20 sand, as utilized in the original test program.

However, NED0-10460A does not reference this alternative material.

Because GE considers llEDO documents to be licensing documents only (nct design documents), GE does not intend to revise NE00-10466A to reflect the substitution of RTV foam material for No. 20 sand.

Where Kaowool had been installed, the design details were reviewed and found acceptable by the NRC.

Fires that are caused by earth-quakes generally involve rupture of flammable liquid and gas ttorage tanks and piping distribution systems.

Since these hazards are not present in nuclear power plant control rooms and since these facilities are design'.d and ceMeutteo te resist and prevent unacceptable damage from earthquakes, earthquake-induced fires are t

not anticipated and are not include' in the design criteria for control rooms.

33

CRGANIZATION: GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA REPORT INSPECTION NO.: 99900403/87-06 RESULTS:

PAGE 8 of 30 The inspectors reconmended that NE00-10466A be revised to reflect the RTV foam material as an acceptable alternative to using sand for applicable plant installat ons.

GE stated their position i

regarding the revision of technical licensing topical reports is:

(1) as a rule, technical topical reports are not revised after they have been approved by the NRC, (2) the topical reports are not intended to provide design requiremants to any design groups within GE, and (3) the topical reports provide an acceptable method for addressing a generic issue or a.way of meeting an NRC staff require-ment based on best available information at the time. GE stated that these licensing topicals can be referenced in specific plant FSARs as a preapproved licensing document.

The concepts and generic designs from NED0-10466A are implemented as specific designs for each PGCC system installed in e nuclear plant.

These designs are described by installation documents provided for each plant and are based upon the design calculations containied in DRF #H13-00071-1 Incax 3, Titled PGCC Fire Stop RTV Silicone Foam Vol. 1 of IV.

Contained within this design file is a Generol Electric Specification; RTV Silicone Rubber Foam Compound No. A1422-5.

l These specifications and design calculations provide the bases for the equivalence of the current designs to the documented fire test.

The thickness of application and specific material to be utilized as well as the suitability of these traterials to perform the functions identified by NED0-10406A were reviewed by the inspectors ard are available within the 'iE Design Record File.

A summary of the main pcints within hED0-10460A is as follows:

l

.Pg 3-4 Fire stops are semi-permanent refractory material (such as r

sand) covered with RTV.

Fire stop f, seal designed to limit air' flow.

.Pg 4-5 Fire tests demonstrate PGCC system ability to maintair, separate redundant Class 1E systems.

.Pg 4-6 Fire stops are easy to install, maintain, and repair.

Refractory naterial covered with FTY linits air flow.

. App F Berkeley Fire Ter.t Feport performed to test the PGC; syst(ri.

Pg 2 Goals of fire test progrm 34 l

ORGANIZATION: GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA REPORT INSPECTION NO.-

99900403/A7-06 RESULTS:

Plcr o nf to Pg 4 Systcra description Pg 7 Fire stopping description F9 63

==

Conclusion:==

PGCC system acceptable.

Fire stops not severely exposed in the 30 minute fire.

RTV silicone fire stops promise to be quite adequate to block fire spread if necessary.

This item is closed.

2.

"?ferred Verification GE Engineering Practices ano Procedures (EP&P) 5.36 Addendon 4, doted December 1975, required thbt a tracking system and status log of deferred verifications be maintaineo.

The inspectors verified during NRC Inspection 86-01 that the first entry was made in the status log for deferred verifications in May 1977. At that time, it could not be determined v:hether verifications had been deferred before May 1977 since the status log did not contain entries of any deferred verifications prior to that date.

During the 86-01 inspection, GE committed to perfonn an extensive review of deferred verifications from inception through May 1977 to positively demonstrate closure of deferred design verifications.

During an NRC review of this effurt during the June 15-18, 1987 inspection, GE had reviewed all 15,300 Engineering Review Memorandums (ERMs) generated from inception to May 1977 to identify ERMs containing a deferred verification statement. As a result, 974 ERMs were identified which affected 3434 design documents.

A computer search of these documents, performed on GE's Engineering Information System (EIS), was then used to identify the current deferred verifica-tion status of the affected documents. GE committed to researching further the status of these documents to verify closure. A DBASE III computer program was used by GE to produce a list of deferred verifications based on criteria established by the NRC during a previous NRC inspection. This criteria was based on safety-related shippable compunents produced by GE NEB 0, San Jose, for use on domestic nuclear power plants. This search produced approximately 150 design documents of which the NRC inspector selected six for further review by GE.

35

ORGANIZATION:

GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA REFORT INSPECTION NO.: 99900403/87-06 RESilLTS:

PAGF 10 nf 10 In a few instances, the six documents selected for review by the NRC related directly to items referenced

'.n Mr. Stokes' sune;ary of Mr. flilam's work record and are associated with Limerick, Susquehanna, and the Shoreham nuclear plant. The six items are:

Item Cocument No.

E R,M, System / Component 1

283X569 BMA 0743 Reactor Vessel Top Guide 2

851E378 CMA 111 Reactor Protection System Elementary Diagram 3

828E375TF AMC 0057 RWCU and Recirculation Bench Board 4

865E152 AMC 0871 RHR/HPCI I Pelay yertical Board 5

237X574Th AMC 0600 HFCI PLV Vertical Board 6

13305530 AMC 0560 RCIC RLV Vertical Board For each of the six documents chosen by the NRC, GE performed a manual search of the entire document history to determine if the opening and closing of all ceferred verifications was accomplished in accordance with establishcd procedures throughout the history of each document.

GE reproduced every revision level of the dccument to identify til ERMs and ECNs which were written against the design ducument.

Each ERM and ECN was then reviewed by CE to identify any deferred verification statement included as part of the Engineering Review Nenorandum or Engineering Change Notice.

The final conclusion concerning these design documents was that all deferred verifications on the six selected documents were found to be completed (closed) as required.

The EIS shows that the six docuoents are complete and verified, and cli 1974-197C deferred verifications on the six selected documents were completed (closed) prior to December 1976.

1he inspector r(viewed edch of the six design documents and ary accompanying ER"s, EChs, DRFs, and other applicable documents to verify that each deferral of t design verification was completed (closed) by a verification statement as part of the ECN.

The completed verificaticn also included a signature and dote in the 36

OhGANIZAT10N: GE NUCLEAR ENERGY SAh JCSE, CAllF0FhiA REP 0f.T INSPECTION NO.:

99900403/87-00 RESULTS:

PAGF 11 of 30 signoif block, "verified by."

In four of the six docun.ents, the Design Record File (CRF) also contained evidence of closure of the deferred design verification in the design verification summary portion.

The six design documents were comprised of an elementary di69 ram, 2 parts list drawings, a connection diagram, and 2 assembly drawings. These design documents averaged 15 revisions during the docurent history and many of the revision levels contained as many as 12 ERMs and ECNs.

The selected sanple of 6 design documents was reviewcd by the inspectors, and no instances were identified where a deferral of a design verification was not eventually completed (closed) as required.

Also, all six were shown as complete and verified on EIS. This item is considercd closed.

3.

ChangestoEngineeringChangesNotices(ECNsl Several entries appear in Mr. Milam's work record and Mr. Stokes' report regarding possible changes madc to ECNs af ter Mr. Milam had sigr.ed the document as the responsible engineer. These items were identified as Stokes report items 1.14, 2.14, 2.28, 2.39, and 5.16.

Inspection finding - A GE review of these items to identify the specific changes noted af ter the sign-offs by Mr. Milam revealed the specific additions or changes to each of the referenced ECN's.

No technical nodification of any document was identified. All of the changes concerned administrative additions or changes where required to keep the docun.ents within the GE document cont'ol system. All of the changes made were easily recognizable with d (ferent printing etc., but were not labeled as to the source of 'ae modification. A memo from Mr. E. R. Welch, Manager Engineering Support (December 1981) identified the requirement that changes to unissued ECNs be initialed and dated by the corrector and the responsible engineer prior to its issue.

Later ECNs provide this additional capability to document control.

The following ECNs/ERMs were noted by Mr. Milam as being changed after his approval and without his knowledge:

ECN/ERM Change / Modification Noted 1)

ECN NJ-13553 - Administrative change on 2nd sheet.

2)

ECH NJ-13555 - Verification statement added - Administration change.

37

ORGANIZATION:

GE NUCLEAR ENERGY SAN JOSE, CAllf0RNIA REPORT INSPECTION NO.: 99900403/87-06 RESULTS:

PArJ 1? nf 7n 3)

ECN NJ-17436 - Editorial change - ro techn' cal impact.

d)

ECN NJ-17441 - The authority block was chaaged by adding "Responsible Engineer" - Administrative change.

5)

ECN NJ-18202 - The authority block was changed by removing the reference to a vendor letter and adding that reference to the narrative description of the change; system MPL block filleo in; distribution identified; The authority block changed by adding "Responsible Engineer" - Administrative changes.

6)

ECN NJ-18?05 - Document title and I?PL number corrected - Administrative change.

7)

ECN NJ-17433 - ECN number corrected (duplicate) and document title changed - Administrative change (new ECN NJ-18E45).

8)

ECN NJ-18?15 - Status of hardware corrected - Administrative change.

9)

ECN NJ-12900 - Revision level of drawing corrected - Administrative change.

10) ECN NJ-18218 - Applicable projects corrected - Administrative chance, 11)

ECN NJ-18235 - Added "Manufacturing Review Required" - Administrative Change.

12) ECN NJ-18236 - Responsible component changed - Administrative change.

13) ECN NJ-18244 - Master Parts List (MPL) number corrected - Acministrative change.

14) ECN NJ-18245 - Corrected hardwaro status - Administrative change.

16) ECN NJ-18249 - MPL number corrected - Administrative change.

16) ECN NJ-19306 - Added "Manufacturing Review Requirea" - Administrative Cnange.

17) ERM AML-2997 - Additional reference edced - Admir,istrative change.

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38

ORGAl'I ZATION: GE HUCLEAR ENERGY SAN JOSE, CALIFORNIA REPORT IhSPECTION N0.: 99900403/67-06 RESULTS:

PAGE 13 of 30 The NRC review did not identify any deviations to the existing GE procedures for docunent control established under GE E0P 55-02, nor any changes that would have affected the technical content of such ECNs/ERMs.

This item is closed.

4.

Stokes Report Section 1.15 "On November E9,1979, Mr. Milat, wrote R. F. Francoise about Category III changes.

The proble was that the draf ting practice when adding parts or groups to dr...ings designates the ERM in the drawing revision block as a Category III change, which is incorrect.

Th F practice derived from the Draf ting Manual, Section 10.1, page 10- whereon an example (Figure 14) illustrated this practice.

Per Mr. Milam's review of E0P 42-6.10, which spells out the responsibilities for ERMs, and E0P SS-E.00, Appendix B, Section B3.1.3, which defines Category III change requirements for ECNs, Mr. Milam was not able to find anything that relates change categories to ERMs, which do not nake changes but rather additions. Mr. Milam requested a revision to the Drafting Manual deletin line 4.6.4.3 and also the ERM reference in Figure 14 (Section 10.1)

Inspection Finding - Mr. Milam was incorrect in his interpretation of the procedures and draf ting manual relative to the exclusivity of the use of ECNs ano ERMs for document application.

Mr. Milan was requesting a revision to the Drafting Manual because he interpreted it as not following the E0P requirements relative to Cateccry III changes and the use of ERMs for such changes.

This request grew out of his observation that CNV Alpha revision drawings were being changed to numerical revision (Rev.1) authorized by Category III EF4 documentation. An ERM is also used to add parts or groups to a document and an ECH is not required. When an ECH is not required, the ERM numbers and the change are entered in the revision area of the document.

This type of revision is a Category III change.

The use of the drawing is protected against unknown changes for production (without-ECN) by control of preliminary l

drawings (alpha revisions) and the release of new group or parts.

l Preliminary (alpha) revision drawings cannot be released for i

production and on Engineering Instruction (EI) is required to release l

new parts or groups.

This item is closed.

l 5.

Stokes Report Section 1.21 "On Friday of the second week of February,1980, Mr. Milam discovered an error in the Corrective Action Request (CAR) programs.

They would not run if there were l

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39

ORGAN 1ZATI0H:

GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA REPORT INSPECTION NO.: 99900403/87-06 RESUtTS:

PAGE 14 of 30 only one CAR.

It appeared that the sort routine "blew up" when it i

tried to sort only one CAn. Mr. Milam entered in his log that he fixed the CAR program during the third weck of February."

Concent:

"There appears to be no progran control, independent verification, or notification of errors to other users; all violete I

one or more sections of 10 CFR 50, Appendix B criteria."

Inspection Finding - This program was only in place for the use of several 6E staff nembers and was a convenience program not a NEB 0 documer.t control or QA program. As such, no official GE controls existed fcr this program.

Violation of 10 CFR 50 Appendix B was not identified.

This iten is closed.

6.

Stokes Report Section 1.23 "During the second week of March,1980, Mr. Nilam completcd a long back-burner job.

he sorted and filed 010 CARS. While doing this, he discovered four (4) white copies (originals):

1.

Three comnitted corrective actions not completed:

SJ-32635 on Kuo Sheng 2 H13-PCO2 written 4-13-78 SJ-32604 on Kuo Sheng 2 M13-P602 written 5-12-78 SJ-3263/. on Kuo Sher:9 2 H13-P60? written 4-13-78 2.

One completed but not returned:

SJ-32240 on Cofrentes H13-P603 written 5 4-70.

Mr. F.ilan was not sure how to handle ther."

Coment:

"Frcn the brief statement, there was sufficient information to indicate a violaticn of 10 CFR 50, Appendix D, Section III, Design Centrol since CARS are engineering documerM tion."

Inspec_ tion Finding - CAks 5a 32634, 32245, 32635, and 32644 for Kuo Sheng Penel ti13-Ft0' were logccd on the Panel Productico Status Log.

J The LChs for torrecting each CAR were lis,ted on the status log ar.d verified as hcVing been u,corporatec. A saraple of other CARS reviewed c'uring the inspection were found to have been closed prior to Mr. Milan 't work recere date. Tne GE document centrol systen operated even thougt the original CARS had beer lef t in Mr. Milam's b6siet. The rurpose of each CAR is to iritiate corrective action via ar ECN.

This iteu is closed.

40

ORGAf112AT 10il: GE NUCLEAR ENERGY SA!! JCSE, CALIFCP.NIA REPORT INSPECTI0ll f,0. -

999004C3/37-06 RESULTS:

PAGF 15 of 30 7.

Stokes Reaort Secticn 2.47 "Reviewing the work record for the third wee ( of July brings a new problem.

The assignment of the same task to niultiple employees."

Ccnn ent:

"This shows that the lack of work control extends on up the ladder of management.

This could be an example of an error by management or, depending on what the ast ignments were, a deliberate dCt.

Example:

Ihe group lead or Mrd ger coJld have a problem which is borderline failing if one engineer does it precisely, it fails, but if it is done by another engineer with less precision, it would pass.

Up to five engineers have been observed working the same mblem with only the calculation which passed used.

The botto n i ne r < rct occuracy but getting the results sought. The most accurate calculations should be used at all times regardless of the results. Otherwise, this should be considered a violation of 10 CFR 50, Appendix B."

Inspection Finding - Calculations could not be found in fir Milam's work record near this tirre period nor could Mr. Stokes' comment be supported for this work effort.

The assignment of more than one person to a particular task is not a violation of any procedure.

This item is closed.

8.

Stokes Report Section 2.48 "In the work record for the third week of July, CNV drawings were being advanced from Revision A to Revision I with Category III changes that in some cases include Design changes (169C9433, H13-P642).

Per E0P, Design changes cannot be Category III. According to a letter from GE's Dave Lee to Mike Hurn on 7-17-80, an Engineering Review Memoranda (ERM) was used to bring the "alpha" document to numeric Revision 1 and that the "Category Ill" changes as authority for changes to drawing is incorrect.

Drawings were still being found with design changes labeled Category Ill during the last week of July 1980. Mr. Milam was not able to obtain any assurance that notations already made in error would be corrected."

Comment:

"This violates 10 CFR 50, Appendix B.Section XVI, Corrective Action."

Inspection Finding - The GE system of documentation control provides a traceable and recorded means of adding engineering changes to a drawing going from alfa to Revision 1 status.

The numeric revision of the drawing changes the status of the design to an authorized 41

ORGANIZATION: GE NUCLEAR ENERGY SAN JOSE, CALIFORNZA REPORT INSPECTION N0.: 99900403/87-06 RESULTS:

PAGE 16 of 30 design that is then under ECN revision control per GE E0P 55.2.

No violation of 10 CFR 50, Appendix B could be identified. This item is closed.

9.

Stokes Report Section 2.52 "On July 28, 1980, Mr. Milam wrote a letter to C. L. Cobler about the revision control of Manufacturing Standard Practices (MSP).

Ref. 1) 159C4279AB, Revision 1, RPS Power Distribution Enclosure A, Connection Diagram, Kuo Sheng 1 and 2.

2)

Inspection Instruction PA-002, Revision 10, Panel Product In-process Inspection.

3)

Inspection Instruction PG-003, Revision 7, PGCC Panel Module Product In-process Inspection.

Mr. Milam said each of the references draws upon the MSP Manual Thus, the MSP becomes a part of the engineering design. However, no reference is made to a particular MSP revision.

For example, 828E342AD Revision 2, Process Radiation Monitoring Instruction Panel Connection Diagram, (Cofrentes) is an instance of non-revision-controlled use of MSP's."

Comment:

"Without a controlled reference number, the references have no traceability.

If used as part of the design documentation for safety-related components, this violated 10 CFR 50, Appendix B."

Inspection Finding - GE maintained controlled processing by the use of a manufacturing log and use of the latest issue of process MSPs for each item.

Traccability was available but inconvenient and no violation of 10 CFR 50, Appendix B was noted. This item is closed.

10.

Stokes Report Section 1.1 "During the second week of May,1978, Mr. Milan recorded that he began ve ification of CNV 1 & 2 and Cofrentes panels H13-P654 and H13-P655.

H13-P654 and H13-P655 are standard Nuclear Reactor Protection System panels for nuclear reactor facilities at Grand Gulf 1 & 2, Black Fox 1 & 2, Clinton 1 & 2, Centrale Nuclear de Valdecaballeros (CNV) 1 & 2, Cofrentes, and Skagit. Mr. Milam encountered a problen with the CNV 1 & 2 and Cofrentes Electrical Device List (EDL). The item numbers did not match those on the standard assembly orawing. Mr. Milam wrote, Note that I did the verification although Russ will sign as verifier, following which I will resolve the verification comments."

42

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l ORGANIZATION:

GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA REPORT INSPECTION NO.: 99900403/87-06 RESULTS:

PAGE 17 of 30 Ccvm.t:

"When one persun does an act which is recorded as having been done by another, such as checking a drawing, several possibilities result.

(1) The document record is false.

(2) The work has not been independently verified, as required by 10 CFR 50, Appendix B."

Inspection Finding - A review of the records pertinent to this item revealed no instances of a lack of procedural compliance for verifica-tion of these panel design documents nor did the records indicate participation by Mr. Milan in the verification of the subject panels. Further, there is no inconsistency with the requirements of 10 CFR 50 Appendix B if the verification activity is delegated to another qualified person providing the verifier is independent of of the design and the original verifier.

The EDL ERMs for CNV and Cofrentes (AMD 1242, 1243, and 1763) associated with H13-P654 and P655 were reviewed.

It was noted that Mr. Milam's name did not appear on any of the ERMs listed. All ERMs except AMC-2588 (issued 3/2/78) were issued in 1977. Mr Thompson's nAme is shown on ERM, AMD 1763 as the design engineer. All EF#s indicate idependent verification. This iten is closed.

11. Stokes Report Section 1.2 "An entry made during the first week of August 1978, identified multiple labeling errors.

In reviewing a draf ting package for H13-P631, Mr. Milam encountered two sets of labeling errors on the isolators (1 set on assembly, 1 set on the connection diagram). With both wrong, he could not use either to correct the other."

Comment:

"The normal progression of draf ting is for the more general type arrangements to be designed, drawn, and checked first, and the more specific and detailed to be drawn using the data from general drawings with additional facts.

Here, it appears that both the order of development between the assembly and connection diagrams were uncontrolled, and neither was accurately checked. This could also be a sign of a et someone else do it" attitude problem among the employees. A violation of 10 CFR 50 exists, since (1) either the procedures do not exist to ensure against this or (2) the procedures are not being followed."

Inspection Finding - Reviewing documents for correctness is a nonnal responsibility for the Responsible Design Engineer. Documentation of activities subsequently indicated that Mr. Milam took action to correct the affected drawings which occurred prior to issuance of the drawing:;. This item is closed.

43

ORGANIZATION: GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA REPORT INSPECTION j

N0.: 99900403/87-06 RESULTS:

PAnF 1R nf 70 J

12.

Stokes Report Section 1.3 "Mr. Milam did some research on document labeling requirements during the third week of August. After reviewing IEEE standard 494-1974, he came to the conclusion that some of GE's drawings were not labeled in accordance with the IEEE standards. Mr. Milam notified the Syracuse office and decided to correct the drawings in his office himself."

Comrrents :

"Without communication and cooperation of management, there is no guarantee that all corrections were made nor corrective training held to prevent the same problem from continuing.

Mr. Milan's actions did not satisfy the requirements of 10 CFR 50, Appendix B.

This reflects a training problem which is obvious in Sections 1.5 and 1.13."

Inspection Finding - Some drawing production was contracted to GE's Syracuse office, but the purchase order did not require adherence to IEEE 494-1974.

It was part e. the job of the Resr,onsible Design Engineer to research the labeling requirements and code applicability in accordance with the governing procedure. This item is closed.

13.

Stokes Report Section 1.4 "Mr. Milan, when reviewing H13-P614, discovered that many devices had not been deleted from the Device List (and EDLs) per earlier instructions from Systems Engineers.

The drawing was per deletico instructions and was not supported by Systems drawings."

Comment:

"Had Mr. Milam been assigned to update the Device List, this entry would not have much importance.

However, he was not.

This indicates that somehow this work was overlooked.

10 CFR 50, Appendix B, Section 111, Design Control prescribes that controls be established to ensure that this does not happen."

Inspection Finding - The drawing errors were identified, checked with the Systems Engineer, cud corrected by direct input to the compJteri;:ed Engineering Information System (EIS).

All documenta-tion was changed as necessary to be consistent. These activities are the responsibility of :he Responsible Design Engineer and were performed by him per GE procedure. This item is closed.

It.

Stokes Report Section 1.5 "A letter to H. H. Hendon en September 19, 1978, concerned the audit of Centro Nuclear de Valdecaballeros (CNV).

Severa' 'f the highli hts listed were of interest, a) The first was 5

thct EGr /E-5.001 Training) was not beino implemented for new employees.

S. Garg (Regnitto, Manager), professed ignorance of procedures due to only having been on the job two months."

44

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PAGE 19 of 30 Conrient:

"There are basically two functions for any auditing department -- (1) to discover problems within the systen; and (2) to resolve the cause of the problems to prevent similar future occurrences.

This entry indicated that the first function was being performed.

The question remaining is, were the problems resolved?

Many entries which followed this one indicate that the problem was not resolved, a violation of 10 CFR 50, Appendix B.Section XVI, Corrective Actioli."

bl "The second was that ECN (NE65571) stated that verification was contained in DRF 921D17 N*3, but the DRF did not contain the verification for this ECN (W. Fraser)."

Inspection Finding - a) This allegation was addressed in NRC Inspec-tion Peport No. 99900403/86-01 and was identified as a nonconformance.

It was closed cut in NRC Inspection Report No. 99900403/87-01 item D.5.

This item is closed, b) Review of the reference DRF showed that verification for ECN NE 65571 is contained therein.

The ECN was prepared on September 25, 1975 and the design verification sheet was signed off on September 26, 1975. This item is closed.

15.

Stokes Report Section 1.10 "A minor problem was noted during the last week of June, 1979.

In reviewing H11-P620 engraving drawing, Mr. Milam discussed some marker plate legends with T. Garg, Systems Engineer for HPCI. Mr. Milam was surprised to learn that T. Garg was not aware that the circuit labels on elementary diagrams are used for marker plate legends."

Consnent:

"This was one of many references to problems with the marker plate legends. Milam's note indicated that Garg previously should have been aware of the use of the circuit labels."

Inspection Finding - It was determined that circuit labels on the system elementary diagrams are used by the panel engineer only as a guide for preparing the panel legend plates and annunciator window legends.

Final marker plate legends are determined by the customer.

The Electrical Systems Engineer is not involved with the preparation of hardware drawings or marker plate legends. This item is closed.

16.

Stokes Report Section 1.11 "During FW7929 (the third week of July, 1979), Mr. Milam completed coment resolution for ERM AMC-3602, with Isolator Terminal Board Assembly and Connection Diagrams.

These i

i 45

ORGAN 12ATION: GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA REPORT INSPECTION N0,-

99900403/87-06 RFSULTS:

PAGE 20 of 30 were sent to Z. Tashjian, Electrical Product Design, for further review.

Virginia Woldow reviewed document quality instead of performing independent design verification."

Connent:

"It is difficult to understand how one can check quality without checking the design, unless quality means the presentation of the work.

In any case, this violated 10 CFR 50, Appendix B."

Inspection Finding - A review of ERM AMC-3602 indicated that adequate and appropriate independent design verification was perfonned following the document quality review. This is evidenced by the comn:ents and responses made to correct deficiencies. This item is closed.

17. Stokes Report Section 1.13 "Entries were raade during the second week of November 1979, when Mr. Milam had attended an EIS users class and an IR class. This was the first record that Mr. Milam had received any training in these two subjects."

Comment:

"Upon consultation, Mr. Milam said that based upon his memory, he had not received any training before this. The lack of any previous training in these areas as well as others is a violaticr.

of 10 CFR 50, Appendix B, Section II."

Inspection Finding - This allegation was addressed in NRC Inspection Report No. 99900403/87-01, item D.5 which closed out nonconformance item 86-01-05, as contained in NRC Inspection Report No. 99900403/

86-01.

This item is closed.

18. Stokes Report Section 1.16 "Mr. Milam made an entry during the third week of December 1979, that Cotrentes H13-F602 became 6 problem on luesday.

Projects wants to ship 'as-is' but panel has many problems."

Comnent:

"Shipping kncwn nonconforming components is a violation of 10 CFR 50, Appendix B, Section XV, Nonconforming Materials, Parts, or Coniponents."

Synopsis:

"On December 14. 1979, the work required by Floor ECNs was not ccmplete but was signed off by the shop. While Mr. Milam was writing this entry on Decenber 18, 1979, the shop was trying tc finish some of the work before the panels were shipped. The work was on Cofrentes H13-P602."

46

ORGANIZATION: GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA REPORT INSPECTION NO.: 99900403/87-06 RESULTS:

PAGE 21 of 30 Corment:

"Signing documentation before completion could result in defective equipment being installed and used.

If these panels were shipped prior to a complete inspection being completed, this was a violation of 10 CFR 50, Appendix B, Section X, Inspection."

l Inspection Findig - Pertinent documentation indicates that the H13-P602 panel for Cofrentes was built in accordance with approved drawings, ECNs and specifications. All floor ECNs were signed off by both Manufacturing and Quality Control.

The shipping records confirm that all work was performed in accordance with established practices, l

i procedures and assigned responsibilities.

However, some of the required equipment was not available at the time of shipment resulting in a "ship short" situation. GE Engineering Operating Procedure (E0P) 55-3.20 defines the controls for shipping panels short of equipny:nt by requiring that a Field Disposition Instruction (FDI) be issued.

In fact, FDI MEEH properly approved by Engineering and the customer was issued.

It was concluded that Mr. Milam's concern, while correct in content, was not valid since proper controls were established for the condition that existed. This item is closed.

19.

Stokes Report Section 1.18 "Weekly report FW8005 to L. C. Wimpee f rom C. W. Hart had an interesting paragraph.

The design for the Hope Creek 1 & 2 Reactor Protection System Vertical Boards H11-P609 and P611 cannot be completed because the elementary drawings continue to change drastically.

Promises by Control and Electrical Engineering are not being kept.

The work-around efforts by both C&I Panel Engineering and Engineering Support have resulted in very much wasted time.

In the interest of supporting projects and the depart-ment, C&I Panel Engineering will continue to complete the design using as many work-around procedures as are possible."

Comment:

"Working around normal operating procedures can only mean that they were violating 10 CFR 50, Appendix B criteria."

Inspection Finding - In the event of design or manufacturing difficulties, a managerial perogative in order to maintain production schedules is to accomplish planned activities in a different sequence or in parallel with other activities. All activities and products arc still subject to procedural, quality control, and regulatory commitments thereby assuring a quality product. This item is closed.

20. Stokes Report Section 1.19 "Mr. Milam wrote ECN NJ-12971 to remove a general interference problem on 163C1122, Conduit Strap, for Joystick Switches during the first week of February, 1980. According 47

ORGANIZATION:

GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA REPORT INSPECTI0t<

11 0. :

99900403/87-06 RESULTS:

PAGF 22 of 30 to his record, this required FDIs for Kuo Sheng 2 and Cofrentes and a very carefully worded disposition because ten other shipped panels are affected."

Comment:

"It is unclear why Mr. Milam had to carefully word the disposition.

It is equally unclear whether the other ten were not fixed, or whether Milam was simply afraid of what management would l

think if they were fixed?"

Inspection Finding - This change was considered to be for product improvement and interchangeability that is governed by GE E0P 55-10.00, "Interchangeability Requirements." All affected plants /

installations must be identified and dispositioned by E0P 55-2.00, "Engineering Change Control." As such, a carefully worded disposition is necessary to clearly define the implementation requirements for the change. Should the customer elect to incorporate the change, an FD1 or FDDR must be prepared as was done for Kuo Sheng 2 and Cofrentes (FDIs MLOG/MEEJ).

It was concluded that Mr. Milam's concern described the normal process for dispositioning a change to a shipped product.

This item is closed.

21.

Stokes Report Section 1.22 "During the last week of February, 1980, Mr. Milan wrote that the drawings for Kuo Sheng C61-P001 did not show the cutcut for the utility outlet. Yet per Pat Falconia, the outlet was installed in a rear post on unit 1 in the field. He stated that he had no design details anc could not make unit 2 like unit 1."

Comment:

"A quality assurance program should ensure that the drawings are accurate representetions of components installed in a nuclear plant, since these records may prove vital during an emergency.

Since the drawing is made from design calculations and sketches which precede thc fabrication of components, the components fabricated should be per the drawing and not vice versa.

Char.ses to fabricated components should not be made without both verifica-tion of the original oesign and changes to the documentation inaicating the changes ar.d their acceptability. The notation above by Mr. Milam's name indicates that these things did not not happen.

j This it, a violation of 10 CFR 50, Appendix B, Section III, Design j

Control, and Section VII, Control of Purchased Material, Equipment, and Services, and others."

Inspection Fir.cing - Corrective action request SJ 53083 was issued and resulted in the acdition of the utility outlet location by ECN NJ 13527.

Subsequently, the cutout type was changed via ECN NJ 48

ORGANIZATION: GE hUCLEAR ENERGY SAN JOSE, CALIFORNIA PEFORT INSPECTION NO.-

999CC403/87-06 RESULTS:

PAGE 23 of 30 13565. This ECN also records the "as built" cutout design and loca-tion for both Unit 1 and 2.

It was concluded that Mr. Milan's concern was incorrect with regard to the documentation of the design details. This item is closed.

22. Stokes Report Sectinn 1.24 "On March 24, 1980, when resolving CAR SJ-54194, Mr. Milam found out that on C61-P001 enclosures the cutouts called out cause interference between marker plates.

Unit 1 engraving drawing had been changed to call for smaller marker plates, but the enclosures were still wrong, as was the Unit 2 engraving drawing.

Fab shop "bootlegged" the cutout; so, marker plates were offset to the side to avoid interference. Mr. Milam did specify rework on these enclosures.

However, there is no written guarantee that the work was done."

Comment:

"Again, this circumvention of procedures is a violation of 10 CFR 50, Appendix B requirements.

See also Section 1.22."

Inspection Finding - Corrective action report SJ 54194 was issued to resolve the problem.

It was then discovered that the marker plate size needed to be revised to accommodate the cutout spacing.

Enclosure drawing 133D9468 was revised to show the enrrect marker plate location with FDis MKCD and MLBL issued to correct the problem.

It was concluded that this was a typical drawing error problem that was corrected by proper procedure. This item is closed.

23.

Stokes Report Section 1.25 "On Thursday, March 27, 1980, Mr. Milam discovered a generic problem on G41-P001 panels. The second row of relays from the bottom had an interference problem. The terminal screws were shorted against welded channels. This was first noticed on Grand Gulf 2 G41-P001, CAR SJ 54138. He observed the problem on panels for Grand Gulf 2, TVA 21, and CNV. He showed the problem to Bob Gordon, QA, 4:15 p.m.

The TVA panels were shipped anyway, Friday."

Comment:

"The fact that QA was notified of this problem not only indicates a violation of 10 CFR 50, but is proof that QA has succumbed to production. This supported Mr. Milam's contention in Section 4.13 that having the QA manager report to Mr. Senn was a problem."

Synopsis:

"Mr. Milam stated that on the last panel he was working on, just prior to his transfer, mylar was used to solve the separa-tion problem when some contacts were shorted against the support angles. He said the mylar was not included on the Parts List."

49

ORGANIZATION: GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA REPORT INSPECTION N0.:

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n PArJ 94 nf 10 Coment:

"This also violated 10 CFR 50, Appendix B criteria."

Inspection Finding - Corrective action report (CAR) SJ 54138 was prepared by the shop to correct the interference problem encountered during the assembly of Grand Gulf panel G41-P001.

Subsequently, ECN flJ13545 was prepared to correct the drawing. A similar problem for TVA panels was resolved later when the same panel for Unit X-22 (cancelled plant) was being fabricated.

CAR SJ56873 was written and ECN NJ20744 was prepared; also Field Disposition Instructions (FDis) were issued to correct the problem on the previously shipped TVA panels.

It was concluded that this was a typical manufacturing and design problem for which procedures for correction were available and implemented.

This item is closed.

24. Stokes Report Section 2.2 "During the first week of April,1980, 1

Mr. Milam was told that non-fungus-proof terminal boards were no I

longer available even though, according to Mr. Milam, many drawings still specified non-fungus-proof terminal boards."

Conment:

"This simple statement indicated that standard details and old drawings, when being issueo for a different job, were not reviewed for accuracy and compliance with that project's requirement for material availability.

Verification should be performed each time a drawing is used, to verify current correctness. Here, this had not been done and over the years material or components which once were available had since stopped being made.

Thus, a noncon-forming part was used. This is a violation of 10 CFR 50, Appendix 8."

Inspection Finding - It was learned that Mr. Milam was misinformed.

General Electric purchased phenolic terminal boards which are fungus resistant from Marathon per drawing 198B6134.

The boards are made non-fungus-proof by the use of a special coating added at extra cost to satisfy customer specifications.

This item is closed.

25.

Stokes Report Section 2.3 "On April 4,1980, a Department memo from C. L. Cobler to Unit 995 on the subject of "ECN Verification State-ment" stated that the best verification statement would list all the documents reviewed thet are associated with the change but that this was impractical with the number of drawings and documents involved in the Floor ECNs."

Convent:

"The likely reason for this being unpractical w6s that this impeded the shipping schedule.

QA should have caught this and stopped this incomplete documentation fron occurring.

The fact 50

ORGANIZATION: GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA RE70RT INSPECTION NO.: 99900403/67-06 RESULTS:

PAGE 25 of 30 that the incomplete documentation took place indicated QA's inability to function. Not only was the incomplete documentation a violatien of 10 CFR 50, but so was QA's inability to function freely."

Inspection Finding - It was determined that the alternate approach of using the Design Record File (DRF) as a compilation of reviewed documents provided an adequate reference for subsequent verification statements.

Therefore, all documentation requirements were still met.

This item is closed.

26. Stokes Report Section 2.6 "In the third week, Mr. Milam discovered that La Salle H13-P645 Panel did not agree with the La Salle Master Parts List (NPL) and that the original Engineering Review Memoranda l

(ERM) for La Salle H13-P644, H13-P645, and others were rejected by the MPL people in January 1979, but issued anyway by Larry Odda.

The ERM listed on the arrangement drawing was the wrong ERM and the parts list description of groups listed group 1 against both La Salle units and did not list group 2.

Corrections attempted by Corky to MPL and Engineering Information System (EIS) and Sam to the Parts List (PL) and ERM number on the arrangement drawing."

Comment:

"These are just two more instances of document or design control problems violating 10 CFR 50, Appendix B."

Inspection Finding - Review of the subject documents indicated that the requirement for listing of parts list group 1 against Unit 1 and group 2 against Unit 2 was met.

Due to schedule commitments, management decided to proceed with the design prior to issuing the ECA and prior to the revision of the MPL adding two panels.

The design was completed and ERM AMJ-046 was signed off in January 1979.

The ECA was subsequently approved on March 28, 1979, and the MPL updated adding the panels. All documentation was determined to be correct. This item is closed.

27.

Stokes Report Section 2.8 "When processing ECN NJ-17584 for the Fuel Pool Cooling and Cleanup System elementary diagrams for Grand Gulf and TVA which was needed along with ECN NJ-13557 to resolve Correction Action Request (CAR) SJ 56233 against TVA 22 G-41-P003, Mr. Milcm discovered that neither ECN was complete. He also found that the panels all come from the same connection diagram but they h:ve different elementary diagrams thus causing many problems when trying to correct. connection diagram problems."

51

ORGANIZATION: GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA REPORT INSPECTION N0.: 90000403/87-06 RESULTS:

PAGF '26 of 30 Comnent:

"This is a problem which is almost impossible to see when developing procedures. Only those using them eventually discover what is wrong.

GE allows one drawing such as a connection diagram to be used for one, two, or nore different projects or plants. This drawing should contain the reference numbers to all associated drawings, i.e., general arrangements, diagrams, details, etc. for which it may be used. As long as the drawing is only used for one or two differer t projects or plants, the problem may not exis~., but at some point, the drawing becomes so complicated that it is virtually impossible for it to be checked without errors being made.

In Mr. Milam's work record, there were many references to drawings which applied to three, four, and rrore facilities. Those entries seemed harmless, but in the context described above could spell i

disaster."

Inspection Finding - Certain par,els previously shipped affected by ECN NJ-13657 had wiring changes incorporated in the field through FDIs.

ECN NJ-17584 was modified by fir. Milam to includa reference to projects and FDIs for shipped units. Mr. Milam's additions to these ECNs were not required since FDIs were listed against hardware ECN NJ-13557.

This item is closed.

28. Stokes Report Section 2.10 "Nearing the week's end, Mr. Milam found out from Bernie that Engineering tells Shop to build things "not per print" with the anticipation of buying it off with an IR (Inspection Report)."

Coninent:

"This results in drawings not being fixed and future problenis in using the same drawings for subsequent units. This practice more than likely stems from the use of deferred verifica-tions. When design verification is put off, the workers beconte less concerned about work quality.

This is because they expect sorreone else to finish what was deferred. A careless attitude develops.

The botton line is that work quality suffers to the point that nothing nay be done correctly.

Bernie's practice, if done without controlling procedures, is a violation of 10 CFR 50, Appendix D criteria, u.d if proceaures are in place which can be used to circumvent normal procedural controls, this is also a violation of 10 CFP E0, Appendix C."

_ Inspection Findino - A review of related documentation could not substantiate Mr. Milam's concerns in this area.

General Electric, however, catagorically claims that it does not fabricate items contrary to drawing requirements.

This iten is closed.

52

ORGANIZATION: GE NUCLEAR EfifRGY SAN JOSE, CALIFORNIA I

REPORT IN5PECTION N0.: 99900403/87-06 RESULTS:

PAGF 77 of 70 29.

Stokes Report Section 2.15 "On April 19, 1980, Mr. Milam was asked to sign-off for separation of Kuo Sheng C61-P001 panel which he had not reviewed. He refused and less than an hsur later he was informed that there were separation problems on t!.e panel.

On a copy of Mr. Milam's work record for R4 8017 was written the following note:

Sam -- The manager of engineering has agreed to perform a separaticn audit on all panels. Please review the separation on > cur responsible panels and sign off the transfer papers requested.

Lee Cobler 4-18-80."

Comment:

"Mr. Milam's efforts did pay off with some action but not by the responsible party.

Here QA should be performing the review, not Engineering."

Inspection Finding - The separation review and signoff was not directed to be done withcut review.

This function was a normal part of Mr. Milam's production responsibilities.

In this particular case, an alternate production engineer performed the function in accordance with management direction and approved procedures. This item is closed.

30.

Stokes Report Section 2.17/2.18 Item 2.17 "During the first week of May, 1980, La Salle Panels H13-PG44 and H13-P645 were shop wired with standard terminal lugs instead of those required on La Salle.

According to his instructions, Mr. Milam attempted to get the customer to accept panels as wired."

Comment:

"In his attempt to get the customer to accept nonconforming panels, Mr. Milam was attempting to avoid rework to the panels, possibly due to a pressing scheduled release date.

He never antici-pated the work that would be required to document this acceptance of nonconforming panels. Had both been weighed, he more than likely would have had the panels fixed instead."

Item 2.18 "During the second week of May,1980, both projects and the customer agreed to accept the panels. However, projects did not want to write a PWA (Project Work Authorization); they only wanted to write a letter. Mr. Milam did no: feel that a letter was adequate to cover a departure from documented project requirements (EWA KA75J19, Revision 0) and QP 10.027. Mr. Milam decided to write an ECN makinn this deviation a temportary substitution."

Comment:

"Per a clarification from Mr. Milam, the meaning of "temporary" as used here is that the change could be done permanently for a specified (temporary) period of time."

53

ORGANIZATION: GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA REPORT INSPECTION NO.-

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PAGE 28 of 30 Inspection Finding - The customer was contacted and agreed to accept the standard terminal lugs. The necessary documentation was prepared to accept the nonconforming panels.

It was concluded that Mr. Milam's concerns were not substantiated since adequate procedures were available and implemented to correct the problem. These items are closed.

31.

Stokes Report Section 2.20 "Mr. Milam talked to John Flaherty on 5-13-80.

Marker plate size on La Salle 1 H13-P644 was changed by Field Disposition Instruction (FDI) but nc one changed the marker plate cutout drawing to accommodate a larger marker plate. Thus, Unit 2 panel was also made using the wrong cutout. John told Mr. Milam that the FDI was incorporated in the field. According to Mr. Milam, this means they drilled new holes in the panel without changing the design drawing. John said it does not matter if the drawings match the panel."

Coment:

"Not only does this indicate a violation of 10 CFR 50, Appendix B, but John's attitude needs correcting."

Inspection Finding - It was determined that ECN's NJ17437 and NJ17438 were processed to correct Parts List 442X206 and drawings 913E800/802.

FDI TDEJ provided corrections for the shipped products and a visual inspection at la Salle confirmed that the ECH and FD1 requirements were implemented.

Corrective action has been taken by GE to incorporate the requirements of ECNs and FDIs into the respective Parts List.

This item is closed.

32. Stokes Report Section 2.21 "During the second week of May, an ECH was prepared to correct the generic design problem on the Hope Creek top covers. Not i ssued.

Inspection Finding - While Mr. Milam indicated that an ECN was prepared, no information could be found on the existance of a generic problem or the issuance of an ECN. With'ut identifying the ECN, Mr. Milam in his docurrented work record reviewed the problem (and the ECN) and concluded that the solution was acceptable. This is also consistent with Mr. Milam's job responsibilities. This item is closed.

33.

Stokes Report Section 2.24 "A recurring problem in Mr. Milan's record is that the wrong engraving drawing has been referenced to a plant. An English engraving drawing has been applied to CNV. The engravings for CNV must be in Spanish."

54 l

ORGANIZATION: GE NUCLEAR ENERGY SAN JOSE, CALIFORNIA REPORT INSPECTION N0.-

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PAGE 29 of 30 l

Comment:

"This problem might seem trivial but could be serious for maintenance and operations personntl.

This surely would be caught at shipment or upon receiving inspection, but maybe not if the people performing those tasks have the same attitude problem as GE's employees in the San Jose office."

Inspection Finding - The process for fabricating marker plates with the correct notations involves referencing the engraving drawings that are in English, issuing an Engineering Work Authorization (EWA) for obtaining a translation in Spanish from the customer, and releasing the correct notatiens to manufacturing for fabrication.

The EWA was issued but was apparently overlooked by Mr. Milam. This item is closed.

34. Stokes Report Section 2.27 "In the first week of June, Mr. Milan was informed that there was a large number of errors on the assembly drawing and parts list for G36-P002.

In view of the number of errors discovered to date in the enclosure drawing, Mr. Milam recommended the drawings be sent back to drafting for a complete recheck."

Coment:

"The drawings do not seem to have been checked properly."

Inspection Finding - Errors in the assembly drawing for control panels were identified on the production floor.

Examples of the errors included noncompatible nuts, screws channels, and f asteners.

Rather than shutting down production for a complete drawing recheck, GE decided to address and correct each problem as it arose consistent with the responsibilities of production engineering associated with manufacturing, testing, and shipping. GE stated that all work was performed in accordance with established procedures and management direction.

It was concluded that Mr. Milam's statement was substantiated and that appropriate corrective action was performed consistent with production floor functions and applicable procedures. This item is closed.

35. Stokes Report Section 2.31 "A new problem recorded was that Manuf acturing Standard Practice (MSP) 14.017 stated that shop-supplied hardware could be substituted for the screws supplied with the switch, but made no distinction between nuclear safety-related switches and others.

La Salle H13-P645 was held up."

Coment:

"This procedural defect is a violation of 10 CFR 50, Appendix B, Section XV, Nonconforming Materials, Parts, or Components. Measures shall be established to control materials, 55

ORGANfZATION: GE NUCLEAR ENERGY SAN J0SE, CAllFORNTA l

REPORT INSPECTION NO.-

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RESULTS:

PAGE 30 of 30 parts, or components which do not conforra to requirements in order to prevent their inadvertent use or installation." See Section 2.86.

Inspection Finding - To accomodate the routing of wiring due to the closeness cf switch spacing ( 4" separation) and the standard shop practice of using ring or spade lugs for terminations, the General Electric production group replaced the saddle clamps provided by the vendor (GE General Purpose Controls) with ring lugs that required longer screws.

However, the vendor indicated that such a change would invalidate the seismic qualification of the switch.

Therefore, GE, under PRC 81-35, seismically tested the revised configuration of the switch and found it to be qualified. All plants to which completed equipment had already been shipped were checked and terminations that differed from the test configuration were made to conform. This item is closed.

56

ORGANIZATION:

NUTHERM INTEP. NATIONAL INCORPORATED POUNT VERNON, ILLINOIS REPORT INSPECTION INSPECTION N0.:

99900779/87-01 DATE:

11/16-20/87 ON SITE HOURS:

172 CORRESPONDENCE ADDRESS: Hutherm International Incorporated ATTN: Mr. L. Hinson, President 501 Scuth lith Street Fount Verncn, Illinois 62864 CRGANIZATIONAL CONTACT: Pr. Ronald heifner - QA Manager TELEPHONE NUMBER:

(618) 244-6000 NUCLEAR INDUSTRY ACTIVITY:

Futhem International Incorporated (NI) fabricates, designs, tests and qualifies electrical devices and control systems for both commercial and military nuclear facilities. Approxirnately 98 percent of NI's business is nuclear related.

il

,]

/// '/

ASSIGNED INSPECTCR:

.o /

/J. Ji Petrosino, Program Development and Reactive / Date Inspec' tion Section (PDRIS)

OTHER INSPECTOR (S):

R. Moist, NRC/hRR T. Tinkel, BNL APPROVED BY:

(d'/ K. Sullivan. BNL <

e i TOh E. T. Baker, Actirs Chief, PDRIS, Vendor Inspection Branch e

INSPECTION BASES AND SCOPE:

A.

BASES: Appendix B to 10 CFR Part 50, and 10 CFR Part 21.

B.

SCOPE:

Follow-up to allegations concerning alleged NI disregard of the NRC QA program requirements in regard to its execution of quality activities.

PLANT SITE APPLICADILITY:

Bellefonte (50-428/439); Braidwood (50-456/457);

Browns Ferry (50-259/260/296); Byron (50-455); Clinton (50-461); Diablo Canyon (50-323/275); Dresden (50-237/249); LaSalle (50-373/374); Limerick (50-352/

353); Nine Mile Point (50-220/410); Oyster Creek (50-219); Peach Bottom (50-277/278); Perry (50-440); Quad Cities (50-254/265); (continued on next page) 57

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PAGE 2 of 24 PLANT SITE APPLICABILITY (continued): RiverBend(50-458);Sequoyah(50-327/

328); Susquehanna (50-387/388); Three Mile Island (50-289/320); Watts Bar (50-390/391); and Zion (50-295/304).

A.

[tplATIONS:

None.

B.

NONCONFORMANCES:

Contrary to Criterion II, "Quality Assurance Program " of Appendix B 1.

to 10 CFR Part 50, and Section 2 of the NI QA manual (QAM):

Section 10 of the QAM does not prohibit a person from inspec-a.

ting their own work; b.

Section 10 of the QAM does not require QA inspection and monitoring activities in NI's equipment testino facility; and NI has not established procedures to control its periodic use c.

of rented measuring and test equipment (M&TE).

2.

Contrary to Criterion III, "Design Control," and Criterion VI, "Document Control," of Appendix B to 10 CFR Part 50; Sections 3.0, 4.0, and 5.0 of NI's QAM and NI QAP #3.0.00:

NI "Qualification Pesults Index" (QRI) forms, which are used as a.

specific haroware instruction riders to NI's gereric functional and test procedures that delineate design parameters for testing, do not indicate that an independent technical review was performed. Additionally, ORI's do not correctly and/or fully translate the design requirements into test parameters and quality standards; b.

Equipment qualification procedures and functional test proce-dures did not indicate that an independent review was performed to verify technical adequacy; and In the pre-1986 time period the engineering manager did not c.

perforn the required design verification activity for several design drawings.

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PAGE 3 of 24 3.

Contrary to Criterion X, "Inspection," of Appendix B to 10 CFR Part 50, NI is not providing adequate QA inspection / verification control of quality related activities as follows:

a.

NI has failed to implement the QA inspection or monitoring program for its equipment testing facility; b.

NI management allows the same technician who performed equipment test activities to inspect his,'her own work and si n the "0A 3

inspector approval" block; and c.

NI management allowed the equipment test facility supervisor to sign the "QA Approval" block for testing results that were performed by his technicians.

4.

Contrary to Criterion XI, "Test Control," of Appendix B to 10 CFR Part 50, NRC observations cf equipment tests and record reviews indicate that NI is neither adequately controlling nor effectively monitoring its safety-related activities that are being performe!

in its equipment test facility.

5.

Contrary to Criterion XVII "0uality Assurance Records," of Appendix B

. to 10 CFR Part 50, adequate records were not in evidence to indicate personnel qualification for several past and present NI employees.

The following HI employee files [ identified by employee initia. '

were found to be either incomplete, incorrect, or indeter71nate as to the relevant experience and/or education: HB, CG, GW, SS, DV, GJ SDJ, LH and PB.

6.

Contrary to Criterion XVIII, "Audits," of 10 CFR Part 50, NI manage-ment allowed the last two QA department annual audits to be led by a QA inspector that has direct QA responsibilities.

The report numbers are QA-86-AE, dated 12/23/86, and QA-85, dated 12/06/85.

C.

OPEN/ UNRESOLVED ITEMS:

Several aspects of the allegation and associated issues were not fully reviewed or evaluated due to the NRC inspection team time constraints.

Therefore, the following items will be classified as open items pending subsequent NRC or vendor actions:

Personnel Qualification (See Section E.4.a)

Correct Performar.ce of Testing (See Section E.4.b)

Radiation Testing Failures of Devices (See Section E.4.c)

Procurement and Receipt Inspection Control (See Section E.4.f)

5) Quality Assurance Procedures / Instructions (See Section E.15) 59

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PAGE 4 of 24 (6) Audits (See Section E.4 9)

(7) 10 CFR Part 21 Potentially Reportable Items (Associated Allegation Aspect)

D.

STATUS OF PREVIOUS INSPECTION FINDINGS:

Heither reviewed nor discussed.

E.

INSPECTION FINDINGS AND OTHER C0FFENTS:

)

1.

Entrance and Exit Meetings The NRC inspection team infortred NI management of the scope of the inspection during the hovember 16, 1987 entrance meeting, and suma-rized the inspection findings, observations, and NRC sthff concerns during the exit meeting on November 20, 1987, 2.

Nutherm Management Cornitments The Nutherm management stait recognized its QA program weaknesses after the NRC inspection team started to identify problem areas, and stopped all of its safety-related activitic! In order to assure its QA program implementation was valid.

Even thcugh NI was not obligated to the Commission to take corrective action until after it received this report, NI voluntarily committed to stop its current work, review its procedures and instructions that control its current ectivities, revise the documents as required, and ensure that the required independent technical adequacy reviews are performed.

The NRC staff believes th6t this approach is appropriate for the circum-starces.

===3.

Background===

This inspection was conducted to examire allegations that have been received by the NRC stcff with respect to the rethod in which NI implerrents its nuclear quality related activitier.

The specific areas of concern that were communicated to NRC staff members include the following:

(a) personnel qualifications, (b) correct perfor-mance of testing and ir, adequate / incomplete procedures, (c) correct-ive action of test deviations, (d) recording of raw test data, (e) adequate translaticn of design bases, (f) procurement / receipt inspection control, and (g) external audits.

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PAGE 5 of 24 4.

Review of Allegations The NRC inspection team observed functional and equipment qualifica-tion tests that were being performed by NI personnel, reviewed NRC licensee purchase order packages, personnel education / qualification records, hardware cualification/ test result records, internal audit reports, program implementing procedures / instructions, design draw-ings/ specifications, and other associated occuments.

The inspectors also conducted interviews with NI personnel at various levels to gain an understanding of how the daily NI quality activities were carried out.

Not every aspect of the allegation was reviewed due to the inspec-tion team time constraints; however, for each aspect of the allega-tion that was reviewed, the results are described below, a.

Personnel Cua11fications:

Allegation Aspect:

(1) Fersonnel with no technical background were given the responsibility to write technical procedures, plans, and reports; (2) would falsify information; and (3) func-tional testing was not performed by qualified personnel.

Inspection Finding:

(1)

It was determined that NI would typically allow non-technical personnel, such as secretaries, to write technical procedures and reports.

It should be noted that this practice is not necessarily incorrect, provided that a technically competent and qualified person performs a review for technical adequacy of the document.

However, it was revealed that N!'s procedures and instructions that were written by non-technical personnel did not always receive technical review and a proval (Nonconfor-rance B.2 was identified in this area.

(2) The efforts of the NRC inspection team did not reveal any evidence that would substantiate this allegation aspect; (3)

Functional testing was not specifically reviewed to deter-mine whether or not qualified personnel were performing the tests in the past; however, current QA inspectors and lab technicians appear to have adequate qualification to perfonn their job functions. A review of NI's personnel files and QA files indicated that, in all cases, NI has 61

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PAGE 6 of 24 not maintained adequate employee files that provide evi-dence of the qualifications of personnel to perform their job activities affecting quality.

(Nonconformance B.5 was identified in this area.) Therefore, this will rt ire a further review.

Based on the above data, this allegation aspect is partially substantiated, and additional reviews will be performed during a future inspection.

b.

Correct Performance of Testing:

Allegation Aspect:

(1) Actual testing sequence of numerous projectswerenotperformedinthecorrectsequence;(2) test-ing prccedures were inadequate; (3) thermal aging tests include i

sloppy record keeping, humidity levels are documented on reports but net in testing records, test time period duration was based on weak link activation energy methodology without determining the lowest activation energy, and devices were not sent out forthermalgravimetricanalysis(TGA);and(4)seismicand design basis event testing contained deviations.

Inspection Finding:

(1) A review of seven equipment qualification packages indicated that their tests were performed in the correct sequence; (2) An overall problem with N!'s technical procedures and instructions was identified.

The inspection revealed that many of the Ni procedures were written by non-technical personnel or by personnel that did not have documented and/or applicable experience to perform their job function.

This methodology would be adequate if the (;A program requirement for an independent "technical review for acequacy" had been perfonced by qualified personnel.

However, the majority of the NI technical procedures were "approved" by the QA manager, instead of the engineering manager as recuired; (3) This area was nnt reviewed in enough detail to determine whether or not NI is maintaining control.

(d)

Ibidem.

Based on the above data, this allegation aspect is sub-stantiated in part.

Additional reviews will be per-formed in this area during a future NRC inspection.

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PAGE 7 of 24 c.

Corrective Action of Test Deviations:

Allegation Aspect:

Radiation testing frilures of devices such as:

thermal cutouts, relays, switches, humidity transmitters, heating elements, auxiliary contacts, and temperature control-1ers have repeatedly failed without NI corrective accion.

Inspection Finding: The radiation testing program was not reviewed; therefore, this allegation aspect is indeterminate and will be reviewed during a future hRC inspection.

d.

Raw Data Recording:

l Allegation Aspect:

"...in many cases raw test data is unavail-l able to review since the raw data was not required for many l

reports and the projects were not reviewed by clientele QA inspectors."

Inspecd on Finding:

The inspection team reviewed the raw data aspect in three areas; thermal aging, design basis events, and rrechanical eging. Seven EQ report packages were reviewed and the raw data [ circular and strip chart recc-ds] were found for all seven. However, in the mechanical aging area, it was determined that the technicians are not recording the raw test data as required after each cycle (Nonconfomance B.4 was identified in this area).

Based on the above data, this allegation aspect is partially substantiated, e.

Inadequate Translation of Design Bases:

Allegation Aspect:

...The testing procedures were inade-quate...some parts that require a three phase load were only tested singe phase... engineering made a guess on how many hours the devices should be thermally aged... fasteners were not torqued...[ personnel] performed tests with no technical testing procedures, EQ plans and reports...."

Inspection Findings:

The NRC inspectors initial observations of NI personnel performing hardware testing activities detemined that technical inconsistencies in its testing methods were obvious.

Discussions with NI Technicians and a review of the procedures and instructions that were established to control the activities determined that an incorrplete and/or inappropriate translation of the design bases was performed in many cases.

The incomplete or inadequate design translation problems were 63

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PAGE 8 of 24 further coarounded because it was also revealed that an indepen-dent review and/or approval was not perfomed in many cases to ensure the technical adequacy of the documents (Nonconformances B.2 and S.4 were identified in these areas).

Based on the above data, this allegation aspect is substan-tiated in part, f.

Procurement and Receipt Inspection Control:

Allegation Aspect:

... Questionable batchlot qualification methodology. basis...with no traceability beck to the manufac-turer's specific batchlot. Some cases of buying from separate distributors at different times and using the same old batchlot numbers...."

This allegation aspect has not been reviewed during this inspec-tion and will be reviewed during a future NRC inspection, g.

Audits:

Allegation Aspect:

... Check into the external audits per-formed by the quality assurance department."

Inspection Finding:

This aspect was not reviewed during this inspection and will be reviewed during a future NRC inspection.

s 5.

Conclusions Concerning Allegations Based on the partial substantiation of several allegation aspects, discussed in E.4 above, the NRC staff concludes that the concerns about the nethod in which NI has controlled its nuclear cuality related activities are valid, and the staff has concerns in regard to the basis of NI's safety-related corrponent Certificates of Con-formance fur previously supplied equipment.

6.

QA Program Establishmeg The inspectors determined by a QA manual (QAM) review that NI's OA program establishrrent in the QAM appears to be basically edequate with the exception of the nonconfomance in 8.1 above.

However, 64

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l

\\

this does not exclude the possibility of other deviations that moy be in NI's QAM, and as discussed with NI management, an NI review should be perfonred.

j Discussion:

At the time of the inspection, it was determined that Class lE activities at Nutherm are categorized inte one of two general creas; production activities, which generally require the assembly of qualified components into a completed unit such as a motor control center, and the second is the equipment qualification test facility where individual components are tested for compliance with environmental or class IE requirements.

Production units were found to be inspected and tested by quality assurance technicians who are independent of organizations responsible for the work and report directly to the quality assurance manager.

Equipment facility test activities, however, were found to be perfonned by technicians who work directly under and report to the test lab manager.

The same technicians are also responsible for the QA verification of work performed (noncon-formance B.1 and B.3 were identified here).

Although this obvious i

I anomaly with Criterion X of Appendix B exists, these activities were found to be in conformance with the Nutherm QAM which does not specifically address itself to nonproduction activities (Non-confonnance 9.1 above),

i 7.

Desion Control The NRC inspection team review of the areas of transferring tech-nical specifications and parameters into procedures and instruc-tions (design translation) and the document review and approval cycle indicated a QA program breakdown in regard to its implementa-tion (Nonconformance B.2 was identified in this area).

Discussion:

Contrary to Criterion 111 of Appendix B to 10 CFR 50, SectT6n 4 of ANSI N45.?-1977, Section 3 and 5 of the NI QAM, Revision 0, dated June 10, 1986 and Section 4 of the NI QAM, Revision 4, dated January 30, 1985, Nutherm procedures and instructions, in general, were found to lack an adequate trans-lation of required test parameters and specifications into the procedures, and often lacked an independent verification of design requirements for adequacy, a.

Due to the wide variations possible for a specific type of curponent, Nutherm has developed the use of Generic Procedures.

The generic test procedures were found to be accompanied by a 65

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PAGE 10 of 24 qualification result index (QRI) instruction fonn which is prepared to define the test requiremeilts and parameters for the specific type of component to be tested.

For the majority of cases reviewed, however, the QRI instruction form was found to lack sufficient detail.

For example, the QRI for Nutherm Project No. BPC-2475 was found to require the cycle aging test to be performed in accordance with Nutherm Procedure No. 9.7.10.10, Revision 3, "Operational Aging of Pushbutton and Selector Switches." This procedure states that the technician is to perfonn the test at a load equivalent to "rated current at nom-iral voltage." However, neither the procedure nor the QRI instruction stated the specific value for rated current.

Based on discussions with hutherm personnel it was identified that the specific values are obtained through the technicians' i

interpretation of the manufacturer's data sheets for the parti-cular device being tested and typically are not delineated or transferred onto the procedures and instructions by the cogni-zant engineers.

Numerous examples were identified, some of which are as follows:

ORI No./Date Prepared by/ Title Approved by/ Title Hardware (1) #1494, 12/85 SDJ/ Design Engineer BE/QA Eng.

Rotary

1. GPU-1759 (Eng.)

Switch (2)(1495,12/85 SDJ/ Design Eng.

BE/QA Eng.

Control

]

I GPU-1759 Switch (3)#1403,10/85 SS/Eng.

LH/Vice President Switch

1. NMP-1841 l

(4)#1467,3/86 PB/EQ Secretary None Selector

1. NMP-1841 (Secy.)

Switch (5) #1759, 1/86 SDJ/ Design Eng.

None Control (GPU-1759 Switch (6)#1514,12/85 GH/ Metallurgist None Rotary

1. GPU-1759 Switch (7) #1529, 1/86 PB/EQ Secy.

None Thyrite fGPU-1897 Protector (8)#1526,1/86 GH/Metellurgist and Fone Current AE/EQ Secy.

Ala.m 66

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PAGE 11 of E4 QRI No./Date Prepared by/ Title Approved by/ Title Hardware (9) #1570, 3/86 S0J/ Design Eng, hone Signal

1. 1759-63 Isolator (10)#1540,1/86 DW/ Technical Asst.

None OSA Relay;

1. FPL-1676 b.

A review of EQ and functional test procedures revealed numerous l

I examples where a technical review of the procedure by an inde-pendent, qualified reviewer, had not been performed.

Specific examples, identified during the review, include the following:

Procedure Title /No.

Prepared by/ Title Approved by/ Title (1) Operations Aging for PB/EQ Secy.

RH/QAManager(Mgr.)

Thyrite Protector,

1. 9.7.10.30, 5/86 (2) Baseline test for PB/EQ Secy.

RH/QA Mgr.

Cycle Duration of Thyrite Protector,

1. 9.7.10.29, 5/86 (3) Baseline Testing of LH/ President RH/QA Mgr.

Solitech Controllers,

1. 9.7.10.22, 12/85 (4) Mechanical Cycle MAM/QA Assistant RH/QA Mgr.

Aging of Airflow Switches, #9.7.10.17, 11/85 (5) Baseline Testing of MAM/QA Assistant RH/QA Mgr.

Disconnect Switches,

1. 9.7.10.6, 11/85 (6) Operational Aging of PB/EQ Secy.

RH/QA Mgr.

Fu:h Button, Selector Switches, #9.7.10.10, 4/86 4

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PAGE 12 of 24 Procedure Title /No.

Prepred by/ Title Approved by/ Title (7) Function Test for AC HB/ Project Mgr.

RH/QA Mgr.

Transformer #9.7.10.39, 9/86 (8) Baseline Testing for SDJ/ Design Eng.

RH/QA Mgr.

Voltage Relays, #9.7.10.26, 1/86 (9) Functional Testing for BE/QA Eng.

RH/0A Mgr.

Temperature Controllers,

1. 7.2.07, 2/85 (10) Functional Testing on SDJ/ Design Eng.

RH/QA Mgr.

Analog and Digital Meters, #7.2.0.6, 3/86 (11) Functional Testing of SDJ/ Design Eng.

RH/QA Mgr.

Solitech Solid State Power Controllers,

1. 7.2.13, 16/86 In addition to the generic functional and environnental qualifi-cation test procedures listed above, the majority of job-i specific QRI instruction forms reviewed were also found to lack an independent technical review.

Examples of these include:

QRI-NTL Number Nutherm Project No.

Rev.

Date 1494 GPU-1759 0

12/23/85 1495 GPU-1759 0

12/23/85 1403 NMP-1041 0

10/29/85 1467 NFP-1841 1

03/05/86 1496 GPU-1759 0

01/20/86 1524 GPU-1759 0

12/26/85 (7) 1529 GPU-1897 0

01/21/86 (8)1526 GPU-017 0

01/07/86 (9) 1570 GPU-1759 0

03/06/86 68

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PAGE 13 of 24 c.

Section 3 of the NI OA Manual, "Design Control " states, in part:

that "The Quality Assurance Manager shall be responsible for verifying the implementation of an effective Design Control program, when applicable...The Project Engineer (s) assigned the responsibility for a specific project shall be accountable for ensuring thnt the requirements of this rection are implemented, when appliceble..." and "lhe Engineering Manager shall be responsible for er.suring that the appropriate technical aspects are performed and that designers, engineers, analysts and verifiers are in conformance with this Design Control program, when applicable."

Sectinn 4 of NI quality assurance procedure (QAP) 3.0.00, Revision 0, 1, and 2, states, in part:

.. 0nce the Project Manager has completed submittal drawings in accordance with Procedures 3.1.00, 3.1.01, 3.1.02, 3.3.00, he shall submit them to the Engineering Manager for approval.

The Engineering Manager shall review for technical acceptability of the design, compliance to the specification and purchase order, and for compliance to procedures." NI QAP 3.0.00, Revision 1, dated July 9, 1986 and QAP 3.0.00, Revision 0, dated February 16, 1985 were reviewed.

These revisions also require the NI Engineering Manager to approve drawings.

Drawings were randomly selected from the master drawing file located in the Engineering Department. A review was performed of selected drawings dating from 1983-1986 to identify engineering personnel involved in preparing, reviewing and approving the drawings.

NI personnel use initials to signify preparation, review ed approval.

Examples of the titles of the below identifi-personnel can be seen in 7.a and 7.b above.

The follrr drawing information was obtained during review of these d,

- -w:

NI NI DRAWN BY ENG.

ENG.

REF. NO.

DWG NO._

(DATE)

REVIEW (DATE)

APPROVAL (DATE)

NPP-2069 7023-56977-53 CTA 5/28/86)

SD 6/24/86)

DES (6/24/86

• GPU-1759 7023-56767-53 JE3 10/29/85) SD 11/11/85)

SS(11/12/85 GPU-1759 7023-56742-53 JEB 10/9/85 SD 10/10/85)

LH (10/10/85 C-1167 7013-55219-53 GAW 12/9/83 SD 1/6/84)

TLS (1/9/84)

• C-1167 7013-55215-53 GAW 12/9/83 SD 4/27/83))

GT(12/13/83) 12/13/83

• BE-1214 5001-54903-43 JDS 4/25/83 SD DJ (4/27/83)

PSE-2185 5001-57018-33 CTA 7/21/86 SD 8/13/86)

DES (8/14/86) 69

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PAGE 14 of 24 NI NI DRAWN BY ENG.

ENG.

REF. NO._

DWG N0.

(DATE)__

REVIEW (DATE)

APPROVAL (DATE)

PGE-1762 4033-57009-33 HEE (7/11/86 HB (7/11/86 DES (7/11/86)

• GPU-2169 7023-56994-33 CTA 8/21/86 SD (8/21/86 HB(8/21/86)

GPU-1759 7023-56749-33 DRM 9/26/85 SD (9/26/85 LH (10/7/85)

GPU-1712 4033-56717-33 DRM 11/21/85) DS (1/31/86 DES (1/31/86)

• GPU-1712 4033-56689-33 DRM (9/26/85)

SS (9/27/85)

SD(10/3/85)

A review of names, init,als and responsibilities of engineering department individuals involved in preparation, review, and approval of the above mentioned drawings indicated that none of the drawings identified above by an asteri:k (*) were approved by the Engineering Department Manager.

The highest level of Engineering Department personnel involved with review and aporoval of these particular drawinas was Project Panager.

So m wing. :n the customer or#r files (NI Reference No.)

were _. mined during a brief gene.al review of purchase order documentation.

The following infornation was obtained during review of these drawings.

NI NI DRAWN BY ENG.

ENG.

REF NO.

DWG NO.

(DATE)

REVIEW (DATE)

APPROVAL (DATE)

SD 2/3/86)

DES (2/4/E6 2/1/86))

FEI 1868 7023-56414-23 DRM 10/7/85 SD 10/9/85)

LH(12/6/85 FE! 1868 7023-56400-33 JEB FEI 1868 7023-56420-33 JEB 10/30/85) SD 10/30/85)

LH (12/6/85 FCI IF68 7023-56401-33 CTA 1/13/86 SD(1/14/86)

DES (1/14/86)

• TVA 1497 1023-55953-33 RW(10/26/84 GJ (10/30/84)

SD (10/30/84)

The drawing identified above by an asterisk (*) was not approved by the Engineering Departinent Manager.

The level of the Engineering Department persennel invohed with the review and approval of this drawing was Project Manager.

The current NI Engineering Departirent Manager (DES) has been in this position since early 1986.

He has a BSEE and is a registered Frofessional Engineer (P.E.).

All drawings selected for review which were issued since he was appointed to this position were found to be approved by him with only the above noted exception. The engineering ruanager indicated thdt in some cases. (re: Dwg. #5023-57401-53 [WPC-2607]), he 70

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PAGE 15 of 24 personally performs the design work if he decides he is most qualified.

In these cases, he typically submits the design drawings for review and approval to his superior, Mr. L. Hinson -

President of N1.

8.

Inspection Activities The NRC inspection team observations, reviews of test result forms, and review of the NI QAM indicates that QA has not been adequately assuring the quality of the work activities in its EQ and functional lab testing facility, v:ith the exception of its annual QA audit of the facility.

In addition, it was determined that NI's lab techni-cians that perform the testing are not prohibited from approving their own test results.

Discussion:

The NRC inspectors decermined by discussions, and obser-vations that QA has not assured that the cuality activities that are performed in the lab facility are verified [where and when necessary] "...by individuals other than those who performed the activity being inspected...," cr, if disadvantageous, provide indirect control by methods such as monitoring processing methcds, equipment, ano personnel (Nonconformance B.1 and 8.3 were identified in this crea).

a.

NI has not implemented its program for the inspection or moni-toring of activities affecting quclity for any of the work activities that are performed in its testing laboratory, except for its annually scheduled QA audit; b.

Discussions with NI lab personnel and a review of several of NI's current typical "functional test result" (FTR) forms that had their associated tests completed and were signed off as being verified by QA, revealed that management is allowing the lab technician that performed the work to sign off the form

[ test results] and use the "0A approval" block for their sig-nature.

Therefore, a review of a completed form indicates that QA review and approval was performed; and c.

Discussions with NI personnel and a review of FTR records indicate that on the previous FTR revision, the test laboratory supervisor was using the "QA approval" block for his signature, indicating that a QA approval had been obtained.

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PAGE 16 of 24 Component qualification testing performed at Nutherm was found to be categorized into either functional testing or mechanical aging testing with test data recorded on specific forms which relate to the type of test conducted.

The previous revision of NI's mechanical aging test result forms were found to require signoff by both the technician and a quality assurance inspector; the current functional test form, hcwever, only requires signoff by a quality assurance inspector.

For all functicnal test forms reviewed, the QA inspector was the same person who performed the work.

Subsequent discussions indicated that the practice of the lab technician who pt.rformed the work signing the block designated as "QA inspector approval" is the allowable NI practice.

In addition, during a review of the pre-vious revision of a mechanical aging test record for NI joo #1841, it was noted that the typical form had two signoff blocks, one for QA approval and one for the lab technician signoff.

In this example, the person who signed in the "QA approval" block was actually the testing lab manager (Nonconformance B.3 was identified in this area).

9.

Test Control The inspection team reviewed documentation packages to determine if EQ reports had been prepared, testing phases were perfcrmed in the proper sequence, raw data was in evidence for design basis event (DBE), thermai aging and mechanical aging, thermal gravimetric analysis (TGA), and other attributes.

In addition, the inspection team reviewed documents and observed testing activities that were currently bcing performed.

E_nvironmental Qualification:

Five purchase orders (P0) were ran-Uomly selected by the NRC inspector frcra NI's computer generated listing.

It was determined by the inspector thet all five P0's were safety-related, had Part 21 and Appendix B to 10 CFR Part 50 imposed. Also, an EQ report had been prepared for each P0.

It was detennined that the EQ testing was for a mild environment for the selected P0's.

The inspector selected seven EQ reports.

Five of the EQ reports represented equipment which was tested for mild environment and the other two were selected from NI's EQ report file and represented equipment which was tested for harsh environ-ment. After reviewing the P0's, it was determined that the requirements of IEEE 323-1974 were imposed.

The testing sequence of ali seven programs were performed in decordance with P0 require-ments.

NI provided the necessary raw data for the desigr basis event and thermal aging.

The inspector did not perform a detailed technical review of this data.

The weak link materials and associated activation energies were discussed in each EQ report.

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PAGE 17 of 24 NI performs numerous generic avalification tests of equipment.

One EQ report (BPC-20?3) was reviewed and evaluated in detail.

The equipment tested was a Microswitch selector switch, a Micro-switch heavy duty oiltight indicator, and Ohmite resistors.

The temperature requirement was 235 degrees F for 30 minutes duration and then 100 degrees F for 3 days.

The requirements were imposed by specification S023-505-5. Devices te::ted were of the same con-struction as those supplied.

The EQ documentation had three evclu-ations recorded which indicated the materials of the tested speci-mens were of the same construction as the supplied equipment.

The testing sequence was verified by reviewing the test specimen work order which showed the tests and dates the tests were performed.

Review of the functional test data sheets showed that the dates tested matched with the dates on the work order.

The functional test data sheets reauired two meciurements to be recorded, however, only one measurement was recorded (Nu;conformance B.4 was identi-fied in this area).

Two TGA reports were reviewed to determine if NI subcontracts material analyses for equipment when materials are not known. As an example, TGA-004, dated February 6, 1986, was an analysis per-fomed by Owens Corning for a General Electric indicator and a Master Specialists illuminated switch.

TGA-001 report dated. Stay 29, 1985 was an analysis performed by Carelco Laboratories for a General Electric motor. NI indicated that prior to 1984 no TGA analysis was performed.

During the NRC inspectors' review of QA procedure 9.7.6,03, dated October 2, 1986, "Procedure for issuing an Equip-cent Qualification Procedure or Report," it was noted that this procedure was not adequate in that it did not reference IEEE-323 (1974), which shows the minimum documentation needed in a EQ file.

The procedure did not show a step by step method of formulating the EQ report.

Other Tests:

In regard to other test lab activities, it was con-cluded that NI is not adequately controlling safety-related acti-vities that are being perfonned in its equipment test lab facility (see nonconformance B.4 above).

The basis for the nonconformance is discussed below:

During a test lab walk through a technician was observed to test undervoltage relays for NI project No. TVA-2605, at a load current of 2.5 times less than the required load.

The technician was asked to demonstrate how the required level of rated load current was determined since this value was not specifically stated within the test instructions or procedure.

The technician submitted for review a copy of the manufacturer's 73

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PAGE 18 of 24 data sheet for the relay under test and stated that the non-inductive load current and voltage used for the test was derived from a table of values located on the data sheet. The inspectors' review of this data sheet, however, found that the voltage and current ratings listed in the table were for inductive load and should have been increased by a factor 2.5 for non-inductive load applications.

In addition, Criterion XI requires, in part, that all testing required be performed in accordance with written test procedures.

Contrary to this, a lab te hnician was observed to perform Environmental Qualification Test Procedure Number 9.7.10.10, Revision 3, dated April 22, 1986 for Nutherm Project Number

)

BPC-2475, without adhering to the procedural requiritments.

Specific examples include the following:

Step 4.1 of the procedure was found to require the use of an adjustable input power source but such a power source was not utilized; Step 4.2 of the procedure specifies the use of an adjustable load bank capable of rated current and nomiral voltage including resistive and inducthe loads. The test setup, however, was found to contain only resistive leading capability in lieu of the required inductive loao; Step 5.1 of the procedure requires one contact of the device under test to be tested at rated inductive lead.

Contrary to this requirement, only a resistive load was being used; and Steps 5.5 through 5.7 of this proc.: dure were found to require the technician to record raw data values of voltage and current following each cycle of the contactor.

Contrary to this state-ment, it was observed during the performance of the tests that the voltace and current values were not being documented as required by the procedure.

The procedure referenced above requires 6000 cycles of the contactor and a record maintained of voltage and current indications following each cycle.

During the performance of the procedure, hcwever, the technician was observed to record voltage and current indica-tions only twice and stated that it was considered a norral practice.

In addition, to a lack of data recording, it was dlso observed that such parameters are not typically monitored during the performance of the test (Nonconformance B.4 was identified in this area).

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PAGE 19 of 24 10.

Personnel Records An attempt was made to review the education and experience of selected NI personnel involved with qu6lity-related activities in regard to writing procedures, instructions, translation of design, technical adequacy review / approval, and similar documents that require varying degrees of various qualifications. Most of the QA training files were reviewed to derive a qualification profile for individuals. However, the information contained in the QA files varied from individual to individual.

For example, resumes were evident for some individuals, while some files had no educational or background information.

However, the payroll files did contain additional educational and background information that were used and considered as QA records by the inspection team.

The NRC inspection team also noted that even when both the payroll and QA files were used some of the combined personnel files were found to be either incomplete, incorrect or indeterminate.

The initials of the appli-cable personnel files that require updating include:

HB, CG, GW, SS, DW, GJ, SDJ, LH, and PB (Nonconformance 8.5 was identified in this area).

11. Audits The inspectica team did not review the area of internal or external audits to any depth.

However, it was noted that NI management has

[

I allowed the use of a QA inspector as the lead auditor on the QA department audits (Nonconformance 8.6 was identified in this area).

12. Measurino and Test Equipment (fi&TE)

The NRC inspectors reviewed the applicable NI QAM section, QAP 12.4.00, "Calibration frequency," and inspected various M&TE equip-ment and logs. Various MATE were found in the lab, receipt inspec-tion, and shop areas.

These M&TE were inspected for M&TE control numbers and for current calibration.

The M&TE inspected are as follows:

NI SER.

CAL.

CAL.

CAL.

ITEM CON. N0.

N0.

DATE DUE BY Laboratory Partlow Temp. Rec.

NI-149 599385 1/14/87 1/14/88 ISL Corp.

Partlow Temp. Rec.

NI-153 6275522 6/17/87 6/17/88 ISL Corp.

beckman Multimeter NI-148 30621038 11/6/87 11/6/88 ISL Corp.

Watlow Temp. Control NI-312 Tag No, 2 9/17/07 9/17/88 GE Evans 75

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PAGE 20 of 24 NI SER.

CAL, CAL.

CAL.

ITEM CON. NO.

N0.

DATE' DUE BY Labcratory Fluke Multimeter NI-302 39410124 2/2/87 2/2/88 ISL Corp.

6/10/87 6/10/88 ISL Corp.

Fluke Ammeter NI-304 11/24/86 11/24/87 GE Evans Torque Wrench NI-322 Multi-Amp Cur. Source NI-330 46381-001/1 5/30/87 5/30/88 Multi-Amp Multi-Amp Ammeter None MNC-2666 7/22/87 7/22/88 Multi-Amp (rented)

Watlow Temp. Cont.

NI-311 Tag No. 3 9/18/87 9/18/88 GE Evans Beckman Multimeter NI-131 10727022 6/22/87 6/22/88 ISL Corp.

Leeds & Northrop Temp.

NI-161 860670871 3/4/67 3/4/88 NI Rec.

-586-08 Analogic Temp. Indic.

NI-157 3002695 6/15/87 6/15/88 ISL Corp.

B&K "0" Scope NI-315 10859 10/5/87 10/88 NI Receipt Inspec.

Continuity Meter NI-159 5080131 9/15/87 9/15/88 ISL Corp.

Fluke Multimeter NI-133 3340111 9/15/87 9/15/88 ISL Corp.

Shop Hypot NI-101 3195 7/24/87 7/24/88 ISL Corp.

Peg-chek NI-103 3351 8/11/67 8/11/88 ISL Corp.

Torque Wrench NI-323 129277 11/26/86 11/26/87 GE Evans The November 17, 1987 M&TE computer printout data base was reviewed and the following were noted:

NI-103. The calibration date on the instrument label was August 11, 1987.

The calibration date in the data base was July 25, 1986.

NI-158. The data base indicates this item is in service and recuired calibration on September 26, 1987.

Upon investigation the item could not be located.

The technician responsible for calibration indicated this item had been removed from service because it failed calibration during its last calibration service.

NI-313.

The data base indicates this item is in service and required calibration on August 28, 1987.

The instrument was located in the shop and was marked with a REFERENCE USE ONLY tag.

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PAGE 21 of 24 NI-315.

The calibration date on the instrument label was October 5, 1987.

The calibration date in the data base was October 1, 1986.

The rented ammeter (Ser. No. MNC-2666) found in the laboratory was not listed in the M&TE computer data base.

Discussions with OA Manager confirmed that NI does not include rented items in the controlled M&TE program, even though this rented M&TE can be used for activities affecting cuality (Nonconformance 8.5 was identified in thisarea).

13..C_ontrol of Purchased Material, Eauipment, and Services The NRC inspectors briefly reviewed a number of NI purchase orders.

However, the brief review only indicated that NI does procure some material directly as nuclear safety-related.

However, most material purchased by NI is comercial grade.

NI then dedicates the commercial grade material for use in nuclear Class 1E safety-related equipment delivered to their customers.

Activities such as receipt inspection, environmental qualification, and final testing of completed assemblies are employed by NI as a basis for commercial grade material dedication. The following are examples of NI purchase orders for commercial orade material:

l P0 P0 Receipt No.

Date Item Grade Insp. Invoked GPU-1759-09 1/23/86 Fuses / Fuse Blocks Com.

Yes GPU-1759-08 10/15/85 Rotary Relay Com.

Yes GPU-1759-07 10/15/85 Resistor Com.

Yes GPU-1759-10 11/8/85 Rotary Pelay Com.

Yes GPU-2573 Yarious Various Com.

Yes Selected NI purchase orders for M&TE calibration services were reviewed.

The vendor's qualification status was ascertained by checking the Approved Vendors List (AVL) in effect at the time of purchase order issue. A summary follows:

P0 P0 Quality 10CFR21 Qualified No.

Date Vendor Req.

Invoked Vendor 3849 10/29/87 Indus. Ser. Lab.

10CFR50 Yes Yes (Per 9/15/87 AVL) 3821 8/25/87 GE Evans 10CFR50 Yes Yes (Per 6/4/87 AVL) 3772 5/7/87 Multi-Amp 10CFR50 Yes Yes(Per3/13/87 AVL) 3693 1/13/87 Indus. Ser. Lab.

10CFR50 Yes Yes (Per 1/13/87 AVL) 77

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PAGE 22 of 24 a.

Purchase order GPU 2573-04 dated May 26, 1987 to Electroswitch was reviewed.

Items 3, 4, and 5 on this purchase order were identified as nuclear safety-related and 10 CFR 50, Appendix B and 10 CFR 21 were both invoked.

The AVL dated May 15, 1987 listed Electroswitch as a commercial rather than a nuclear vendor. Additional review and discussion with the QA Manager revealed documentation that showed that the QA Manager upgraded Electroswitch to e ruclear supplier on May 7,1987, but the status code on the AVL was not changed. As a matter of additional note, the status code had not Seen changed as of the September 1987 issue of the AVL.

b.

A number of receipt inspection records for selected purchase orders were reviewed.

In general, receipt inspection consists of identifying material through verification of part numbers, verifying proper shipping quantity, and checking for visual shipping damage. The following receipt inspection documenta-tien was reviewed:

Receipt P0 Item Inspection Date No.

No.

Accepted Inspector Note 6/4/87 2573/03 2

Yes AE 1

6/4/87 2537-03 6

No AE 2

6/10/87 2573-03 Various Yes TW 6/11/87 2573-03 1

Yes TW 6/24/87 2573-03 5

Yes TW 9/14/87 2573-03 1

Yes AE Not Recorded 1759-07 1.2 Yes Not Recorded 1

1/29/86 1759-05 1.2 Yes AE 1

Note 1:

The NI control number of the M&TE used for testing was not recorded.

Note 2-Nonconformance report NCR 1263 dated 6/4/87 was issued.

14.

Purchase Order Review The hRC inspection team reviewed several NRC Licensee Purchas Orders (P0's), that resulted in several of the discussed nont formances.

Sone of the NRC Licensee P0's that were reviewed ii 78

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PAGE 23 of 24 NI Customer 10CFR21 Ref. No.

Plant P0(Date)

Item Invoked Appendix B Environment CVI 1599 Palo Verde P5-5665 Remote Yes 10CFR50 (2/4/85)

Panel Control CTI 1037 Palo Verde 11212 Elec-Yes 10CFR50 (3/11/81) trical Heater CPL 1654 Brunswick 28030 Contact Yes N45.2 Mild (3/18/85)

Relay 1654 Brunswick 28031 Micro-Yes N45.2 Mild (3/18/85) switch FLUOR 1868 San Onofre 441207-6 AC/DC Yes Mild (1)

-0002-01 Dist.

(11/13/85) Panels COMSIP 1449 Hope Creek E-7922-181 Humidity Yes 10CFR50 Mild (1)

(8/10/84)

Element N45.2 TVA 1497 Watts Bar 85K-9-835985 Duct (10/12/84) Heater Buffalo Forge 1618 River 72726 Power Yes Bend (1)

(1/28/85)

Stud Block 1665 Bellefonte 85K5-8369 Bailey Yes Yes 48 (5/7/85) Power Supply 1670 Fitzp6 trick 6N029 Panels Yes Yes (12/9/85) w/BKR's 79

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PAGE 24 of 2d NI Customer 10CFR21 Ref. No.

Plant P0 (Date)

Item Invoked Appendix B Environment 1699 Peach Bottom DE 348898 2 pole /

Yes Yes (4/29/85) 600 y BKR's 1701 Millstone 843241 CH Relays Yes Yes (6/20/85)

XFMP, TB's BKR's 1708 Oyster Creek PP 0288839 Starter Yes Yes (8/29/85)

Unit, Motor CKT BKR 15.

Quality Assurance Procedures The QA procedures and instructions were not reviewed within the scope of this inspection.

However, the NRC inspector perceives that problems will also exist in this area based on the above findings. Therefore, this area will be reviewed during a future NRC inspection.

i F.

NI PERSONNEL CONTACTED:

• W. Eckert, Chairman

• L. Hinson, President

• T. Stomberski, Vice President

• D. Stephens, Engineering Marager

• I. Gunin, EQ Manager

• C, Baker, Lab Manager

• R. Heifner, QA Manager

• M. A. McCann, QA

• S. Wadt, Admin.

• J. Winder, Engineering T. Johnson, EQ S. DeJournett, Engineering G. Pierce, Lab B. Revelle, QA A. Evrard, QA R. Winder, Lab

• Attended exit sceting.

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PULLMAN CONSTRUCTION INDUSTRIES, INCORPORATED CHlCAG0, ILLINCIS REPORT INSPECTION IhSPECTION N0.: 99900853/87-01 OATES:

12/07-11/87 nN_U TF Hn m - 7?

CCRRESPONDENCE ADDRESS:

Pullman Construction Industries, Incorporated ATTN: Mr. L. H. Goldwyn President 1400 East 97th Place Chicago, Illir.cis 60628 j

ORGANIZATIONAL CONTACT: Mr. M. A. Jarigese, QA Manager TELEPHONE NUMBER:

(312) 374-6700 NUCLEAR INDUSTRY ACTIVITY: Pullman Construction Industries, Incorporated (PCI) is currently providing 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Underwriter's Laboratory (UL) rated curtain type fire dampers to NRC 11censees that are enveloped within an outer flange assembly that provides for thermal expansion and seismic considerations.

1. N l

l/

\\

ASSIGNED INSPECTOR: /

L., //[ M 4 6

/#h / Y #?6

'~ioseph J. Petrosino, Program Development and Date Reactive Inspection Section (PDRIS)

OTHER INSPECTOR (S): Anthony Fresco, Brookhaven National Laboratory Joseph Ulie*, USNRC - Re ion III APPROVED BY:

[

7 Edward T. Baker, Acting Chief, PDRIS, Vendor Inspection Date Branch

• Inspection participation on December 7, 1987.

INSPECTION BASES AND SCOPE:

A.

BASES: Appendix B to 10 CFR Part 50 and 10 CFR Part 21.

B.

SCOPE:

The inspection was performed to evaluate the validity and basis of closure tests conducted by PCI of their curtain type fire damper assemblies performed under simulated in-plant air flow conditions and to review the implementation of its QA program in certain areas.

PLANT SITE APPLICABILITY:

Generic, all plants.

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PAGE 2 of 10 A.

VIOLATIONS:

None.

B.

NONCONFORMANCES:

1.

Contrary to Criterion I, "Organization," Criterion II, "Quality Assurance Program' and Criterion III, "Design Control," of Appendix B to 10 CFR Part 50 and Sections II and III of PCI's quality assurance (QA) manual, PCI has failed to establish an adequate QA program to control its safety-related activities in the areas of design and QA monitoring for fire dampers.

Examples of this include:

a.

Objective evidence could not be provided that would indicate that PCI had performed an independent review for technical adequacy of its fire camper "suitability for application" evaluations that it has performed for several NRC licensees; b.

Objective evidence could not be provided that would indicate that PCI had performed an independent review for technical adequacy of its fire damper performance qualification curves; c.

PCI has not adequately delineated the authority and duties of its design engineers who perform activities affecting the functions of safety-related fire dampers nor has it delineated associated monitoring duties for its QA engineers; d.

PCI has not established adequate engineering procedures or instructions to provide ccrtrol over its activitics affecting its safety-related fire dampers; e.

It could not be deternined by a review of objective evidence whether or not PCI has correctly translated its in-hcuse fire damper raw test data into PCI design paraneters or instructions; and f.

Calculations and engineering determinations for safety-related fire dampers are typically perforced by PCI engineering personnel without benefit of procedures, instructions or CA monitoring activities.

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PAGE 3 of 10 C.

OPEN/ UNRESOLVED ITEMS:

The below listed issues were discussed during the inspection but were not fully and or adequately resolved.

Therefore, the followir.g items are classified as open itens and will be discussed during a future NRC inspec-tion and/or subsequently by NRC staff:

1.

Testing of fire dampers for closure under air flow at ambient tempera-tures, in lieu of elevated temperatures which would be experienced during a fire; 2.

Industry curtain type fire dampers (CTFD) closure certification of dampers based on testing of one CTFD of a given size and design; 3.

The validity of the fire rating of fire dampers by the Underwriters Laboratory (UL), that are based on "tne assumption that air conditioning and ventilating systems a;e automatically shutdown at the start of a fire" (emphasis added), when automatic shutdown is not designed into the system and does not occur; and 4.

The applicability to nuclear power plants of the Air Movement and Control Association, Incorporated (AMCA) standard number 500, which is the only industry standard that delineates methods to test dampers under flow ano defines a fire damper as a device that closes under flow and ef fectively stops air flow. This differs from the UL basis for its fire duration rating.

D.

STATUS OF PREVIOUS INSPECTION FINDINGS:

None.

E.

OTHER FINDINGS OR COMMENTS:

1.

Entrance and Exit Meetings The NRC inspector informed PCI management representatives of the scope of the inspection during the entrance meeting on December 7, 1987, and sunnarized the inspection findings and observations during the exit meet:ag on December 11, 1987.

===2.

Background===

This inspection was conducted to review and evaluate CTFD closure tests that were conducted by PCI under simulated in-plant air flow 83

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PAGE 4 of 10 conditions. The testing was performed by PCI at the American Warming and Ventilating (AMV) Company test laboratory in Toledo, Ohio. The tests were conducted to qualify a PCI modification to the typical CTFD design configuration and were all witnessed and documented by AMV personnel. The PCI modification did not change the CTFD blade or frame; instead the change consisted of adding corregated steel segments to the CTFD outside frame members, and then fabricating a dual element sleeve around the existing CTFD.

These changes add rigidity and provide room for thermal expansion for the resulting PCI fire damper assembly unit.

It was also observed that the PCI fire damper assembly would appear to better withstand rough handing and installation at the end user's facility. The PCI assembly would also appear to remove the pcssibility of typical warping problems during installation caused by differer,t welding techniques used during plant installation.

However, PCI does not manufacture the actual fire damper itself, but procures safety-related CTFD's from Air Balance Ir.corporated (ABI),

Westfield, Massachusetts. The ABI fire damper is then encapsulated, without change to its blades or track, into the forementioned sleeve assembly which is designed by PCI to prcvide for both thermal expansion and seismically induced forces.

2.

Purpose of Testing PCI conducted its cwn testing to determir,e the product operability parameters and effects of the folicwing variab!es on damper closure under simulated in-plant air flow:

Upstream duct length and configuration Downstream duct length and contfiguration Damper ar.d flow orientation Static pressure Velocity Positive and negative pressures Damper size Height-to-width ratio Spring tension l

l An important aspect of the PCI testing appears to be its establith-l nient of the maximum total (or stagnation) air flow pressure at which a given damper will successfully close.

It appears that the 84

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PAGE 5 ef 10 operability of a ClFD is more dependent on the total differential pressure at the damper and less dependent on the associated air flow rates or velocities.

PCI has put its accumulated raw test data into a report which then is used to certify closure of its damper under different in-plant duct work configurations and air pressure variables.

3.

PCI Test Data Usage and Creditability usage:

Four reports were presented in a binder entitled, "Pullman Construction Industries, Incorporated - Fire Damper Data." The first is a seismic qualification of HVAC fire dampers for the Shearon Harris plant dated December 20, 1984.

The second describes CTFD closure tests for ASI Model 319 horizontal and vertical fire dampers dated hovember 1984 The third is a report by Underwriters Laboratories on a fire test of a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> CTFD by Air Balance utilizing an alternate method of installation.

This report is dated December 5, 1984.

Lastly, the fourth is a test report by American Warming and Ventilating, Incorporated, on a PCI CTFD assembly, dated July 18, 1985.

After discussions with PCI personnel as to the specific purpose of the reports, it was determined that:

Fnclosure No. 1 is intended to be a seismic qualification of PC1's fire damper assembly.

Enclosure N ). 2 is intended to show erratic performance of existing da npers not supplied by PCI.

Enclosure do. 4 is intended to show actual air flow closure performance of PCI's fire damper assembly.

Enclosure No. 3 was not separately reviewed but the report itself states that it is intended to determine whether an alternate installa-tion of a fire damper provides resistance to the passage of fire for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and to the effects of a hose stream following fire exposure as specified by the UL Standard 555 fire endurance and hose stream tests.

It was noted that each report was presented as a separate entity with no explanation of:

The intended purpose or conclusions to be drawn from each report.

The interrelationship between each test rer. ult.

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PAGF 6 of 10 At this point in the inspection, the NRC team queried PCI personnel in regard to a possible misuse or m!sinterpretation of its test reports by either hRC staff or by licensees that do not have PCI's CTFD assemblies installed.

PCI's management comitted to address this NRC staff concern by generating an explanation of the intention of each document and the relationship of each document to the others in a foreword signed by appropriate PCI personnel. This will address the NRC concern over possible misuse of the test repor.s to justify fire damper installa-tions which were not supplied by PCI.

Credibility: After discussion with PCI personnel, it was determined that the appropriate test data for review by the team was enclosure No. 4, which PCI utilizes as a source of test data to establish the sizing and configuration of its damper assemblies.

PCI also had supported a fifth test report performed by American Warming and Ventilating, Incorporated of PCI's horizontal and vertical CTFD's. This report, dated Septer.ber 26, 1985, is consid' red prcprietary by PCI.

It includes test data performed for R&D purposes where important modifications were made to the dampers and then the modified dampers were subsequently tested for closure under air flow.

PCI personnel stated that its only bearing on their product is that, in some cases, the data can be used to verify that more than one sample of a given size and configuration was tested to support the data in the July 18, 1985 report.

It was revealed that only one sample of a given size and type damper is required to be tested by industry standards such as UL 555 and AMCA 500. There are no specific industry requirements concerning the minimum number of samples to be tested.

For the tests reported on in the July 18, 1985 report, it was noted that the test sampics were PCI standard stock items.

In view of these factors, it was decided that only the July 18, 1905 PCI report would be considered in evaluating the acceptability of PCI's testing program for closure under air flow since its results are the most comprehensive.

Since the intended purpose of the PCI testing is to establish the maximum total (or stagnation) pressure at which a given damper will successfully close, PCI utilized the results of its testing to plot i

l a curve of damper area versus the pressure upstreau of the damper 86 J

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PAGE 7 of 10 upon complete closure (i.e., the total pressure since the flow velocity is zero after closure with the supply fan still operating).

This curve, together with a PC-based computer program which PCI uses as an encineering tool to develop the damper specifications, is then interpreted as establishing an upper limit on the total pressure at which a given darrper will clcse, thereby leading to certification by i

l PCI that a damper design established by tL informational input will l

close under air flow conditions specified by a licensee.

An independent plot of the curve was performed by the BNL representa-tive and the data points properly coincided with the PCI plot.

However, this should not be construed to be a verification of the technical adequacy of the PCI test data or design program.

==

Conclusion:==

PCI's fire damper testing methodology appears to have simulated some worst-case in-plant air flow conditions, using a sound engineering rationale, and its test results and conclusions appear to be crediable based on the inspectors review. The PCI documented testing observations also appear to have identified cre of the two major reasons why CTFDs industry wide are experiencing difficulties or failures during air flow closure attempts. One reason appears to be the total differential pressure as discussed above. The other reason appears to be that the CTFD's were not originally designed to close under air flow ccnditions, because licensees did not specify air-flow closure requirements on purchase orders in the time frame of 1970-1980.

4.

Industry Standards Currently, the only industry standard that has been typically imposed by NRC licensees in regard to CTFD is the Underwriters Laboratory (UL) Standard number 555 (UL-555), "Fire Dampers and Ceiling Dampers." The UL-555 intent states, in part:

"It is the intent that tests made in conformity with the test nethods described herein will develop data to enable regulatory authorities to determine the acceptability of fire damper assemblies for use in locations where fire resistance of a specified duration is required."

The UL-555 scope also states, in part, that:

"Closing reliability of fire dampers is evaluated on the basis that air-conditioning and ventilating systems are _ automatically shut down when a fire occurs as described in the various provisions of the Standard for the Installation of Air-Conditioning and Ventilating Systems, NFPA Number 90A, Therfore, the ratings are applicable to fire dampers and ceiling dampers installed in systems v:here air movement is effectively stopped at the start of a fire" [ emphasis added.]

87

ORGANIZATION:

PULLMAN CONSTRUCTION INDUSTRIES, INCORPORATED CHICAG0, ILLIN0IS REPORT INSPECTION NO.: 99900853/87-01 RESULTS:

PAGF A of JL The Air Movement and Control Association, Incorporated (AMCA)

Standard Number 500 (AMCA-500) was adopted by the AMCA membership on October 18, 1973 and its stated purpose is to establish uniform test methods for relating pressure drop and/or water penetration to air flow rate for louvers, dampers and shutters. The AMCA-500 definition of a fire damper is:

"...a device arranged to interrupt air flow automatically through part of an air system so as to restrict the I

passage of flame, and is installed...to close automatically in the event of a fire in order to maintain the integrity of the fire-rated separation."

The AMCA-500 standard has several different test setup configurations for fire damper closure tests under air flow, with guidelines for associated instrumentation types / configurations for the monitoring and recording of flow rates, static pressure drops, seating torque, and other related measurements.

PCI conducted its own testing program primarily because UL-555 is the only industry standard that has been historically imposed and applicable to fire darnpers and ceiling dampers installed in systems withir, nuclear plants.

Since the PCI product is intended to provide advantages in installations required to meet seismic requirements, seismic tests were conducted in addition to closure under flow and structural integrity tests.

Furthermore, the tests provided R&D information on the usefulness of various modifications intended to improve perforatnce leading to justification of the final product design. Ultimately, the test results establish a data base which PCI utilizes to specify product size and configuration (i.e., whether a single or n.ultiple damper is reovired for a given size and flow condi tion).

5.

Design Centrol The NRC inspectors reviewed documents and conducted discussions to determine whether PCI's design control activities in regard to its fire damper assenbly test program are adequately controlled.

It was determined that PCI has not adequately controlled its design ontrol program as evidenced by the lack of objective evidence to

'sure that the quality activities were satisfactorily acccoplished.

The basis and the results of the fire damper test program were reviewed to ascertain several aspects which included:

l (1) Whether the test result raw data [ design basis) was correctly translated into design requirenents;

{

I 88

ORGANIZATION: PULLMAN CONSTRUCTION INDUSTRIES, INCORPORATED CHICAG0, ILLIN0IS REPORT INSPECTION N0.-

99900853/87-01 RESULTS:

PAGF n nf in l

(E) Whether the design requirements were indepcndently reviewed and approved for technical adequacy; and (3) To verify whether an acequate review for "suitability of applica-tion" is being performed.

The review of the above three areas revealed that PCI had failed to adequately control its cuality program activities. As an example, during a review of PCI's "suitability of application" design activities, it was revealed that the same PCI engineer typically performs all the design activities and approves his own work. The design activities include a review of a licensee's ventilation system design parameters and characterics for determining the particular size, type, and/or configuration of the PCI CTFD that will adequately perform their function within the licensee's ventilation system.

Hence, PCI cannot show that someone other than the person who performed the design activities independently reviewed it for technical adequacy (Nonconformance B.1 was identified in this area).

6.

Procedures Instructions The NRC inspection team reviewed the PCI quality assurance manual design section and attempted to review PCI's design procedures and instructions.

It was revealed that PCI has failed to adequately establish its design control program and failed to establish proce-dures that would assure that important engineering and quality assurance activities have been satisfactorily accomplished in regard to its fire damper assembly program (Nonconformance B.1 was identified in this area).

The NRC review revealed that PCI has failed to adequately implement its QA program requirements in regard to assuring that quality activities are adequately controlled by procedures and instructions that delineate the applicable criteria.

As an example, the NRC inspector requested to review the applicable procedure that delineates how the PCI engineer performs his review for suitability of component application at a licensee facility.

The inspectur was shown how each review is performed; hcwever, PCI did not have a procedure that defined the job functions and responsibilities of the engineer for this aspect.

It was also revealed that each of the other engineering activity areas in regard to PCI's test program were not delineated in regard to its QA program control and monitoring.

89

ORGAN 1ZATION:

PULLMAN CONSTRUCTION 1NDUSTRIES, INCORPORATED CHICAG0, ILLIN0IS REPORT INSPECTION NO.: 99900853/87-01 RESULTS:

PAGF 10 nf 10 7.

PCI Personnel Contacted

• L. H. Goldwyn, President

• J. P. Goldwyn, Sr. Vice President

• R. F. Cekanor, Executive Vice President

• G.

Zielinski, Vice President

• M. E. Saucier, Project Manager

• T.

I. Stewart, '

ject Manager

• M. A. Jarigese, QA Manager

• K. Schaeflein, Shop Manager

• S. Szykowny, Welding Foreman

• R. Baker, Welder

• Attended exit meeting.

i 90

SELECTED INFORMATION NOTICES

UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C.

20555 January 11, 1988 NRC INFORMATION NOTICE NO. 86-81, SUPPLEMENT 1:

BROKEN EXTERNAL CLOSURE SPRINGS ON ATWOOD & MORRILL MAIN STEAM ISOLATION VALVES Addressees:

All holders of operating licenses or construction permits for nuclear power reacto s.

Purpose:

This notice is provided as a supplement to Informatien Notice No. 86-81 to alert recipients of information about additional closure spring failures on main steam isolation valves (MSIVs) that have been reported at Fermi Unit 2.

The MSIVs were manufactured by Atwood & Morrill and the springs that failed were manufactured by Duer Spring and Manufacturing.

It is expected that recipients will review the inforirstion for applicability to their facilities and consider actions, as appropriate, to avoid similar problems.

However, suggestions contained in this information notice do not constitute NRC require-ments; therefore, no specific action or written response is required.

Background:

At Fermi 2, in May 1986, the licensee observed that four external closure springs (the inner members of the affected pairs of external springs) were broken into several pieces.

The licensee analyzed two of the broken springs and determined the failure to be quench cracking caused by the heat treatment process during manufacturing.

The spring manufacturer, Duer Spring :-d Manufacturing, also performed laboratory metallographic examination of the railed springs and veri-fied that the failure was the result of quench cracking.

Both Atwood & Morrill and Duer recommended that all the external closing springs on all MSIVs be cleaned and subjected to magnetic particle testing at the earliest opoortunity.

Duer provided an inspection procedure.

In addition, General Electric (GE) issued Service Information Letter No. 422, dated July 18, 1986, recommending a visual inspection and in some cases load tests.

It also stated that springs passing the 105 percent load compression test would be expected to provide satisfactory service at normal loads.

The broken springs at Fermi were replaced with springs that had been compressed to 1 5 percent normal load.

No magnetic particle irspection was conducted at that time.

8801050394 91

IN 86-81, Supplement 1 January 11, 1988 Page 2 of 3 Description of Circumstances:

At Fermi 2, on March 21, 1987, the licensee found two additional broken springs of the same type and location.

These springs had been inspected according to the GE recommendations.

The licensee attributed the failures to pre-existing seams and quench cracks (induced during heat treatment) that had propagated as brittle fracture failures.

The spring material, which is the same as that of the springs that had previously failed, had undergone temper embrittlement.

The licensee plans to replace all currently installed valve springs in early 1988.

After heat treatment, but before they are painted, the replacement springs will be subjected to the following tests:

magnetic particle, full compression (108 percent of the maximum operating stress), and spring constant.

Discussion:

The licensee performed a fracture mechanics analysis of the data, taking into account temper embrittlement.

Their calculated value for critical crack size in temper embrittled material under 100 percent maximum operating stress is 0.03 inch.

Their estimate for critical crack size ird temper embrittled material under 105 percent maximum operating stress is 0.027 inch.

These results are essentially equal.

Measurements of critical crack size on 7 actual fractures varied from 0.044 inch to 0.077 inch.

These crack sizes are smaller than those visually detectable, particularly when the surface is coated.

Consequently, the licensee concluded that the 105 percent load test followed by a visual examination would not suffice for separating defective springs.

It should be noted that a previous analysis contracted for by the licensee did not take into account temper embrittlement effects and did lead to an earlier conclusion that the 105 percent load test would suffice.

The failures of springs which had been subjected to 105 percent load tests and visual inspection appear to confirm the licensee's conclusion that the root cause is temper embrittlement, quench cracking, and surface imperfections.

The licensee's corrective action of conducting full compression testing followed by magnetic particle testing thus appears to be appropriate.

Springs from Duer Spring and Manufacturing have been supplied to both pressurized-water reactor and boiling-water reactor nuclear power plants on MSIVs manufactured by Atwood & Morrill.

Atwood & Morrill has identified those plants that would be affected by the problem.

This list is attached.

l l

1 92

IN 86-81, Supplement 1 January 11, 1988 Page 3 of 3 No specific action nr written response is required by this information notice.

If you have any questions about this matter, please contact the technical contact listed below or the Regional Administrator of the appropriate regional office.

h arles E. Rossi, Director l

Dl vision of Operational Events Assessment Office of Nuclear Reactor Regulation Technical Contacts:

R. L. Cilimberg, NRR (301) 492-9656 Vern Hodge, NRR (301) 492-8196 Attachments:

1.

List of Affected Plants as Identified by Atwood & Morrill 2.

List of Recently Issued NRC Information Notices 93

IN 86-81, Supplement 1 January 11, 1988 Page 1 of 2 LIST OF AFFECTED PLANTS AS IDENTIFIED BY ATWOOD & MORRILL CO., INC.

The following utilities have main steam isolation valves furnished with exter-nal closing springs manufactured by Duer Spring & Manufacturing Co.:

PRESSURIZEO-WATER REACTORS:

Utility Plants Quantity Valve Size Duke Power Catawba 1, 2 8

34-inch Duke Power McGuire 1, 2 8

32-inch Duke Power Perkins 1*, 2*, & 3*

12 32-inch Duke Power Cherokee 1*, 2*, & 3*

12 32-inch Houston Lighting South Texas 8

32-inch i

& Power Project 1, 2 l

South Carolina Virgil Summer 1 3

32-inch Elec. & Gas TVA Bellefonte 1, 2 8

32-inch TVA Sequoyah 1, 2 8

32-inch TVA Watts Bar 1, 2 8

32-inch

• Unit cancelled l

l f

94

IN 86-81, Supplement 1 January 11, 1988 Page 2 of 2 BOILING-WATER REACTOR,S_:

Utility Plants Quantity Valve Size Boston Edison Pilgrim 1 8

20-inch Cleveland Elec.

Ferry 1, 2 16 26-inch Illuminating Detroit Edison Enrico Fermi 2 8

26-inch Georgia Power.

Hatch 1 8

24-inch Gulf States River Bend 1 8

24-inch Houston Lighting Allens Creek 1*

8 24-inch

& Power Illinois Power Clinton 1 8

24-inch Jersey Central Oyster Creek 1 4

24-inch l

Power & Light I

l Mississippi Power Grand Gulf 1, 2 16 28-inch

& Light Niagara Mohawk Nine Mile Point 1 2

24-inch j

Power Northern States Monticello 8

18-inch Power Pennsylvania Susquehanna 1, 2 16 26-inch Power & Light Philadelphia Limerick 1, 2 16 26-inch Electric Philadelphia Peach Bottom 2, 3 16 26-inch Electric

)

Public Service Hope Creek 1, 2*

16 26-inch l

Elec. & Gas l

TVA Browns Ferry 1, 2, & 3 24 26-inch l

TVA Hartsville Al*, A2,*

32 26-inch l

81*, 82*

l TVA Phipps Bend 1*, 2*

16 26-inch

• Unit cancelled l

l I

95 L

IN 86-81, Supplement 1 January 11, 1988 Page 1 of 1 LIST OF RECENTLY ISSUE 0 NRC INFORMATION NOTICES Information Date of Notice No.

Subject Issuance Issued to 87-67 Lessons Learned from 12/31/87 All holders of OLs Regional Inspections of or cps for nuclear Licensee Actions in Response power reactors, to IE Bulletin 80-11 87-66 Inappropriate Application 12/31/87 All holders of OLs of Commercial-Grade or cps for nuclear Components power reactors.

87-28, Air Systems Problems at 12/28/87 All holders of OLs i

Supp. 1 U.S. Light Water Reactors or cps for nuclear l

power reactors.

87-65 Plant Operation Beyond 12/23/87 All holders of Ols Analyzed Conditions or cps for nuclear power reactors.

87-64 Conviction for Falsification 12/22/87 All nuclear power of Security Training Records reactor facilities holding an OL or CP and all major fuel facility licensees.

87-35, Reactor Trip Breaker 12/16/87 All holders of OLs Supp. 1 Westinghouse Model 05-416, or cps for nuclear Failed to Open on Manual power reactors.

Initiation From the Control Room 87-63 Inadequate Net Positive 12/9/87 All holders of OLs Suction Head in Low Pressure or cps for nuclear Safety Systems power reactors.

87-62 Mechanical Failure of 12/8/87 All holders of OLs Indicating-Type Fuses or cps for nuclear power reactors.

87-61 Failure of Westinghouse 12/7/87 All holders of OLs W-2-Type Circuit Breaker or cps for nuclear Cell Switches.

power reactors.

OL = Operating License CP = Construction Permit 96 l

UNITED STATES NUCLEAR REGULATORY COP 911SSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C.

20555 February 5,1988 NRC INFORMATION NOTICE NO. 88-04:

INADEQUATE QUALIFICATION AND DOCUMENTATION OF FIRE BARRIER PENETRATION SEALS Addressees:

All holders of operating licenses or construction permits for nuclear power reactors.

Purpose:

This notice is to alert addressees that some installed fire barrier Pietration seal designs may not be adequately qualified for the design rating of the penetrated fire barriers.

It is expected that recipients will review this information for applicability and consider actions, if appropriate, to pre:lude a similar problem and correct existing problems at their facilities.

However, suggestions contained in this information notice do not constitute new NRC requirements; therefore, no specific action or written response is required.

Description of Circumstances:

The NRC has been reviewing fire barrier penetration seal designs installed in several nuclear power plants. The reviews focused on whether the installed configuration was qualified by adequate testing and documentation.

The current NRC review was prompted by reports, inspection findings, allegations, and other information that indicated the possibility that NRC requirements for fire barrier p'netration seals were not being net in all aspects. The review included: eva.gations of fire barrier penetration seal specifications and pro-cedures developed by licensees, licensee agents, and licensee contractors; evaluations of various fire barrier penetration seal tests and test data; and inspections of various fire barrier penetratien seal designs and installations.

The types of concerns identified to date and mentioned below are related to weak-nesses in the implementation of NRC requirements and guidelines as related to fire barrier penetration seal design qualification.

The staff identified instances where installed fire barrier penetration real designs could not be verified as qualified for the design rating of the pene-trated fire barrier.

In some cases, test oualification documentation was not 8802020228 97

---e,,

e v--n a

,,.,.,.,-,,--,,,--,-,,,-,,,,.-,n

IN 88-04 February 5,1988 Page 2 of 3 available, in other cases, qualification test documentation was available but incomplete or inadequate because all qualification requirements had not been satisfied or the installed seal design configuration cr design parameters were significantly' different from the tested seal.

The NRC review also has identified a current practice that can affect the qualification status of installed seals. Pla~nt modifications are being made that require running new cable and conduits through existing penetration seals.

These modifications are generally being made without an associated technical review to ensure that the resulting penetration seal design configuration or design parameters are consistent with those validated by initial qualification tests. Over a period of time, numerous minor modifications to the same area

-could cumulatively result in a degraded fire barrier rating.

Discussion:

NRC requirements and guidelines for fire barrier penetration seals are con-tained in various documents, including Appendix R to 10 CFR 50, Appendix A to Branch Technical Position (BTP) APCSB 9.5-1, "Guidelines for Fire Protec-tion for Nuclear Power Plants Docketed Prior to July 1, 1976," and NUREG-0800, Standard Review Plan. The extent to which these requirements or guidelines are applicable to a specific plant depends on plant age, connitments established by the licensee in developing the fire protection plan, the staff safety evalua-tion reports (SERs) and supplements, and the license conditions pertaining to fire protection.

The goal is to provide a fire barrier penetrdtion seal that will remain in place and retain its integrity when subjected to an exposure fire, and subse-quently, a fire suppressing agent. This will provide reasonable assurance that the effects of a fire are limited to discrete fire areas and that rne division of safe-shutdown-related systems will remain free of fire damage.

A number of licensees have conducted a comprehensive assessment of the adequacy of in-plant fire barrier penetration seals. Their efforts began by determining which specific NRC guidelines / requirements apply and which specific connitments were made to respond to those guidelines or requirements. Typ'cally, in-plant seal assemblies were surveyed to catalogue the various types of existing seal configurations.

Finally, the documentation was analyzed to confirm that in plant designs were fully qualified by a fire test and were installed in a proper manner.

If these efforts revealed instances where seals were not installed where re-quired, were not installed properly, or were not qualified by a standard fire test, then the licensees have considered the :eals degraded and have in.ple-mented corrpensatory measures, such as fire watch patrols, per the appropriate technical specifications or administrctive procedurer. These measures remain in force pending final resolutiors of the issue.

FintTresolutionmayinclude replacing existing penetrati,on seels with fully qualibei seals, qualifying '

in-plant seal asserrblies by Wpslemntah fire tests, ard justifyino Jn-plara i

l configuration by fire hazards / safe shuttoxn analysis. 1 g

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IN 88-04 February 5,1988 Page 3 of 3 Appendix A contains a summary of various technical considerations that have been used for evaluating the qualification adequacy of fire barrier penetration seal designs and installations, associated testing, and test data.

No specific action or written response is required by this information notice.

If you have questions about this matter, please contact one of the technical contacts listed below or the Regional Administrator of the appropriate NRC regional office.

8er f Y

( m Uarles E. Rossi, Director Division of Operational Events Assessment Office of Nuclear Reactor Regulation Technical Contact (s):

Dennis Kubicki, NRR (301) 492-0825 Joseph Petrosino, NRR (301) 492-0979 Attachments:

1.

Appendix A - Summary of Existing Staff Guidance Related to Fire Barrier Penetration Seals 2.

List of Recently Issued NRC Information Notices i

s

(

8 99

IN 88-04 February 5, 1988 Page 1 of 3 APPENDIX A

SUMMARY

OF EXISTING STAFF GUIDANCE RELATED TO FIRE BARRIER PENETRATION SEALS A.

General Considerations Concerning the Use of Test Results To Qualify fire Barrier Penetration Seal Designs The (fire barrier seal) test specimen shall be truly representative of the construction for which classification is desired, as to materials, work-manship, and details such as-dimensions of parts, and shall be built under conditions representative of those obtaining as practically applied in building construction and operation. The physical properties of the materialt and ingredients used in the test specimen shall be determined 2

and recorded 8.

Seal Acceptance Criteria 1.

Thefireresistanceratingofthepenetrationsellshouldbgequiva-lent to the rating of the barrier in which it is installed 2.

The fire resistance rating of the penetration seal should be deter-mined by a standard fire test (i.e., ASTM E-814, ASTM E-119 or IEEE-634) 4 3.

The test should be conducted by an independent, recognized testing authority.

The tested assembly should be representative of in-plant assemblies. The exposure fire should correspond to at least the time-temperature curve of ASTM E-119. Thermocouples should be positioned at representative locations on the cold side of the tested assembly (including the intarface of seal material and through penetrations).

The cold-side temperature should not exceed 250'F above ambient during the test or 325'F maximum, although higher temperatures at through penetrations are permitted when justified in terms of cable insulation ignitability. There should be no burn-through of the seal during the test, nor the passage of hot gases sufficient to ignite cotton waste material. The assently should withstand tge effects of a hose stream, as stipulated in the standard test method The seals should be installed by qualified individuals.6 4.

5.

Appropyiate quality assurance / quality control methods should be in force.

6.

Fire barrier penetrations that must maintain environmental isolation or pressure differentials should be qualgfied by test to maintain the barrier integrity under such conditions l

l 100 1

IN 88-04 February 5, 1988 Page 2 of 3 9

C.

Hose Stream Testing 1.

Hose s ueam testing is a requirement for all fire barrier penetration seal qualification testing, regardless of whether the penetration l

seal is for a wall or a ceiling or a floor.

2.

Hose stream testing should be performed on tested specimens that have successfully withstood the fire endurance test requirements.

3.

The hose stream shall be delivered in one of the following ways:

a 1-1/2-inch nozzle set at a discharge angle of 30* with a nozzle pressure of 75 psi and a minimum discharge of 75 gpm with the tip of the nozzle a maximum of 5 ft from the exposed face; a 1-1/2-inch nozzle set at a discharge angle of 15* with a nozzle pressure of 75 psi and a minimum discharge of 75 gpm with the tip of the nozzle a maximum of 10 ft from the exposed face; a 2-1/2-inch national stan-dard playpipe equipped with 1-1/8-inch tip, nozzle pressure of 30 psi, located 20 ft from the exposed face.

4 The duration of the hose stream test should meet the minimum require-ments specified in ASTM E-119 for firt barriers. During hose stream testing, the fire barrier penetration seal should remain intact and should not allow a projection of water beyond the unexposed surface.

10 D.

Deviations Deviations from NRC requirernents or accepted industry standards for fire barrier penetration seals should be technically substantiated as part of the review and approval of the fire protection plan or in other separate formal correspondence. Supplemental guidance is provided in Generic Letter 86-10.

References 1.

"The design of fire barriers for horizontal and vertical cable trays should, as a minimum, meet the requirements of ASTM E-119, Fire Test ofBuildingConstructionandMaterials, including]thehosestreamtest."

[Section D.3.(d) of Appendix A to BTP APCSB 9.5-1.

"Penetration seal designs shall utilize only noncombustible materials and shall be qualified by tests that are comparable to tests used to rate fire barriers." (Section Ill.M of Appendix R to 10 CFR Part 50).

"The penetratior qualification tests should use the time temperature exposure curve specified by ASTM E-119."

(Section C.5.a. of BTP CHEB 9.5-1).

101

IN 88-04 February 5, 1988 Page 3 of 3 2.

ASTM E-119, "Fire Test of Building Construction and Materials."

3.

Section 0.3.(d) of Appendix A to BTP APCSB 9.5-1.

4.

.Section Ill.M. of Appendix R to 10 CFR Part 50, 5.-

ASTM E-119, "Fire Test of Building Construction and Materials."

6.

Section C of Appendix A to BTP APCSB 9.5-1, Section C.4 of BTP CMEB 9.5-1.

7.

Ibidem.

8.

Section C.5.a.(3) of BTP CMEB 9.5 1.

i 9.

Section Ill.M of Appendix R to 10 CFR Part 50, Section C.5.a of BTP CMEB 9.5-1.

10. Generic Letter 86-10.

?

j' r

I 102 i

UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR RFACTOR REGULATION WASHINGTON, D.C.

20555 February 29, 1988 INFORMATION NOTICE NO. 88-06:

FOREIGN OBJECTS IN STEAM GENERATORS Addressees:

All holders of operating licenses or construction permits for pressurized water reactors (PWRs).

Purpose:

This information notice is being provided to alert addressees to a potentially generic problec with foreign objects on the secondary side of steam generators in PWRs and the potential for failure of steam generator tubes as a result of frettino.

It is expected that recipients will review the information for ap-plicability to their facilities and consider actions, as appropriate, to pre-ciude similar problems from occurring at their facilities. However, suggestions contained in this informatiun notice do not constitute NRC reouirements; there-fore, no specific action or written response is required.

Description of Circumstances:

On February 3, 1988, during the first refueling outage at Catawba 2, the licensee submitted a report to the NRC, as required under 10 CFR 50.7?, indicating that foreign metal objects in two of four steam generators had caused loss of 50% of the wall thickness for one tube in each of these steam generators. One of the objects was a steel block 1-1/2 inches x 2 inches x 3 inches and the other was a steel sliver approximately 1/4 inch x 3/8 inch x 8 inches.

Both objects were located on the tube sheet. On the basis of a subsequent evaluation, the licensee concluded that it was unlikely that these objects had caused damage.

However,.

the licensee found three jacking studs 2 to 3 inches long by 2-1/4 inches in diameter on the tube sheet in one of the other two steam generators. One of these studs did cause significant damage to a tube in that steam generator.

Tne licensee also found small gauge wire, a nail, a 6 inch piece of welding rod, and weiding slag in the steam generators.

The licensee was able to remove all of the foreign objects and debris except for the piece of welding rod.

The licensee did eddy current testing of 100% of the pertcheral tubes in all of the steam generators for C'.tawba 2 and found that plugging of seven tubes was neces-sary. Of the seven tubes plugged, two had visible damage to the exterior surface near the tube sheet.

The maximum defect indicated by eddy current testing was 77%.

The licensee's records indicate that all of the steam generators were carefully inspected for foreign ob.iects after fabrication of the steam generators was com-pleted and prior to operation of the reactor at power.

Nothing of significance 8802230035 103

IN 88-06 February 29, 1988 Page 2 of 2 was found at that time. The foreign objects and debris found recently in the steam generators may have accumulated in the top works of the steam generators and may have been washed down to the tube sheet af ter preservice inspections.

Discussion:

As described in NUREG/CR-0718. "Steam Generator Tube Integrity Program Phase i Report " tests for certain steam genera"rs indicate that tube failure can be expected during nonnal operation if the tube wall thickness is reduced by 85%

to 90% and if the defect is 1-inch long or longer. Likewise, during a steam line break accident, failure can be expected if wall thickness is reduced by 75% to 80%.

Information Notice 83-24 "Loose Parts in the Secondary Side of Steam Generators at Pressurized Water Reactors," addressed events involving foreign ob.iects and loose parts on the secondary side of steam generators at PWRs, including the events at Ginna and Prairie Island I where tube ruptures were related to the presence of loose parts.

Generic letter 85-02, "Staff Recomended Actions Steruning from NRC Integrated Program for Resolution of Unresolved Safety Issues Regarding Steam Generator Tube Integrity," requested that PWR licensees perfonn visual inspections on steam generator secondary sides in the vicinity of the tube sheet along the entire periphery of the tube bundle and the tube lane to identify any foreign objects and damage to the external surfaces of tubes.

For licensees with operating licenses, these inspections were to be perfonned during the next planned outage for eddy current testing and after any secondary side repairs or modifications to steam generator internals were completed. For applicants for operating licenses, these inspections were to be performed as part of the preservice inspection program.

The licensee for Catawba 2 complied with this request as an applicant for an operating license. The preservice inspections would not detect foreign objects in the top works of the steam generators that could be washed down to the tube sheets during the first fuel cycle.

No specific action or writter response is required by this information notice.

If you have questions about this matter, please contact the technical contact listed below or the Regional Administrator of the appropriate regional office.

Charles E. Rossi, Director Division of Operational Events Assessment Office of Nuclear Reactor Regulation Technical

Contact:

Roger iloodruff, NRR (301) 492-1180

Attachment:

List of Recently issued NRC Information Notices 104

Attachment IN 88-06 February 29, 1988 Page 1 of 1 LIST OF RECENTLY ISSUED NRC INFORMATION NOTICES Information Date of Notice No.

Subject

-Issuance Issued to 88-05 Fire in Annunciator Control 2/11/ 88 All holders of Ols Cabinets or cps for nuclear power reactors.

88-04 Inadequate Oualification 2/5/88 All holders of Ols and Documentation of Fire or cps for nuclear Barrier Penetration Seals power reactors.

P8-03 Cracks in Shroud Support 2/2/88 Ali holders of Ols Access Hole Cover Welds r cps for BWRs.

88-02 Lost or Stolen Gauges 2/2/88 All NRO licensees authorized to possess gauges under a specific or general license.

88-01 Safety injection Pipe 1/27/88 All holders of OLs Failure or cps for nuclear power reactors.

86-81, Broken External Closure 1/11/88 All holders of OLs Supp. 1 Sprinos on Atwood & Morrill or cps for nuclear Main Steam Isolation Valves power reactors.

87-67 Lessons Learned from 12/31/87 All holders of Ols Regional Inspections of or cps for nuclear Licensee Actions in Response power reactors.

to IE Bulletin 80-11 87-66 Inappropriate Application 12/31/87 All holders of Ots of Connercial-Grade or cps for nuclear Components power reactors.

87-28, Air Systems Problems at I?/28/87 All holders of OLs Supp. 1 U.S. Light Water Reactors or cps for nuclear power reactors.

87-65 Plant Operation Beyond 12/23/87 All holders of OLs Analyzed Conditions or cps for nuclear power reactors.

OL = Operating License CP = Construction Permit l

105

_. _ _ _. - ~._.

UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C.

20555 March 18, 1988 NRC INFORMATION NOTICE NO 88-09:

REDL'CED RELIABILITY OF STEAN-DRIVEN AUXILIARY FEE 0 WATER PUMPS CAUSED BY INSTABILITY OF WOODWARD PG-PL TYPE I

GOVERNORS l

Addressees:

All holders of operating licenses or construction permits for nuclear power reactors.

Purpose:

This information notice is provided to alert addressees to continuing problems affecting the reliability of steam-driven auxiliary feedwater pumps caused by instability problems with Woodward governors.

It is expected that recipi-ents will review the information for applicability to their facilities and consider actions, if appropriate, to avoid similar problems. However, sugges-tions contained in this infonnation notice do not constitute NRC requirements; therefore, no specific action or written response is required.

Description of Circumstances:

The steam-driven auxiliary feedwater pumps at Calvert Cliffs are powered by Terry steam turbines (GS-2N) with Woodward PG-PL type governors.

Before July 23, 1987, periodi:: surveillance testing of the steam-driven auxiliary feedwater pumps (AFWPs) was preceded by a warmup of the turbines before initiation of the quick startup tests. On July 23, 1987, during a loss-of-offsite-power event, the number 11 AFWP on Unit I tripped on its initial demand as a result of turbine overspeed. To ensure that future pericole (monthly) surveillance testing of the turbine-driven AFWPs would be con-ducted under more realistic conditions, the test procedures were modified to require quick starts from cold conditions. During subsecuent tests in July throuah October 1987, a number of trips of the stean-driven AFWPs oc-curred at Calvert Cliffs. On July 30, 1987, during rapid cold startup testing, both of the Unit 2 steam-driven pumps tripped. On September 26, 1987, the number 11 AFWP on Unit 1 began oscillating after the initial startup attempt and subsequently tripped on overspeed. On October 23, 1987, the number I?

AFWP on Unit 1 tripped on overspeed.

The **censee conducted an intensive testing and troubleshooting program to detemine the causes of the failures. During these tests, a n'imber of test 8?"3140284 107

IN 88-09 March 18, 1986 Page 2 of 3 failures were experienced because of turbine governor oscillation and overspeed.

The most frequent failure sequences were either rapid initial acceleration of the turbine to the overspeed trip point, oi large undamped speed oscillations that increased in magnitude to the overspeed trip point. Less freauently, trips occurred when the mechanical latch mechanism holding the trip valve open (which appeared to be overly sensitivel tripped. Subsequent attempts to test the pumps imediately after initial steam-driven AFWP failures were normally successful.

Discussion:

Several factors were identified which appear to have contributed to the reduced reliability of.the AFWPs. These include:

1.

Use of governor buffer springs of less than optimal stiffness, resulting in the inability of the governor to dampen out upset conditions. One of the installed governors and all three of the spare governors had buffer springs of a lower stiffness than that listed in the procurement specifi-cation on file at the Woodward company.

2.

Excessive condensate trapped in the steam supply lines, resulting in governor valve damage, governor 1.inkage damage, and throttle control instabilities as slugs of water hit the governor valve and turbine wheel.

3.

Improperly adjusted and degraded governor linkage, resulting in excessive linkage play.

4 Governor valve binding, resulting in governor actuator over-reaction to small feedback signals.

5, A failed governor on the Unit 2, number 22 AFWP.

6.

Damaged and misaligned overspeed trip mechanisms, resulting in oversensitivity to vibration, jarrino, and waterhamer.

The licensee implemented several corrective actions and plans additional upgrades.

These are described below.

1.

Stiffer buffer springs were installed in the governors of all AFWPs to increase control system dampening at the expense of increased control system response time.

2.

Upgrading of both the procedures and the systems was initiated, which included more thorcugh drain procedures and drain lineup verification.

The interval for manually drainino the steamlines and turbine casings was decreased from ever; 8 to every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Additional manual drains were installed in the system low points to eliminate water from the steamlines.

3.

Various parts of the covernor valves, governor linkages, and trip linkages were overhauled, adjusted, and replaced. Trip linkages associated with 108

TN 88-09 March 18, 1988 l

Page 3 of 3 the overspeed mechanisms and the trip throttle valves were ad,iusted to increase trip latch engagerient and thereby reduce sensitivity to physical shock.

For some parts, such as linkage plates (cams), it was necessary for the utility to obtain the special materials involved and fabricate replacement parts in house.

4 Further steamline drain improvements are being evaluated.

The Calvert Cliffs problems highlight the importance of optimally sizino buffer springs, since the single, most effective short-tem corrective action appeared to be installation of the stiffer buffer springs.

However, changes in spring stiffness for the purpose of improving stability can adversely affect other governor response characteristics. Therefore, the selection of optimal spring stiffness should be carefully considered.

In addition, it is important to ensure maintenance of proper spring stiffness following initial detemination of optimal stiffness.

In the case of Calvert Cliffs, the addition of stiffer springs appeared to provide an extra margin of stability.

This temporarily compensated for other auxiliary feedwater system deficiencies that also re-quired correction.

Reliability problems were much more evident when the auxiliary feedwater pumps were periodically tested using quick starts from cold conditions.

This demon-strates the importance of surveillance testing which, in so far as practical, duplicates the service conditions that would exist if the equipment were called on to operate.

Infomation Notice 86-14. "PWR Auxiliary Feedwater Pump Turbine Control Prob-lems," and P6-14, Supplement 1, "Overspeed Trips of AFW, HPCI, and RCIC Tur-bines," discuss problems closely related to those discussed in this information notice.

No specific action or written response is required by this infonration notice.

If you have any Questions about this matter, please contact the technical contact listed below or the Reoional Administrator of the appropriate NRC regional office, dZ%

h r es ossi Tirector Division of Operational Events Assessment Office of Nuclear Reactor Regulation Technical

Contact:

D. Limroth, RI (215) 337-5121

Attachment:

List of Recently Issued NRC Infomation Notices 109

UNITED STATES NUCLEAR REGULATORY COPNISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C.

20555 i

April 18, 1988 NRC INFORMATION NOTICE NO. 88-14: POTENTIAL PROBLEMS WITH ELECTRICAL RELAYS Addressees:

All holders of operating licenses or construction permits for nuclear power.

reactors.

Purpose:

This infomation notice is being provided to alert recipients to potential problems involving HFA, PVD 21B, PVD 210, and HGA relays manufactured by General Elet.tric Company (GE), as described below. Although some of these problems are several years old, they are included because utilities are still identifying problems discovered while implementing advice issued by GE in the early 1980s. The specific problem with the HFA relays concerns mechanical binding in the relay caused by incorrect location of a stop tab that is welded to the armature.

The problems with the PVD 21B, PVD 210, and HGA concern their seismic capability.

It is expected that licensees will review this infonnation for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in this information notice do not constitute NRC requirements; therefore, no specific action or written response is required.

Description of Circumstances:

l a.

HFA Relays In June 1986 Duane Arnold Energy Center reported that HFA relays were not resetting. GE determined, after investigation, that mechanical bindino in the relays prevented the nonnally closed contacts from making contact when the relays were de-energized. This was caused by the incorrect location of a stop tab that is welded to the amature.

The incorrect location, com-bined with possible movement of the magnetic assembly, causes the armature binding.

In July 1986, TVA experienced similar problers.

On July 21, 1986, the GE Peter and Control Business Division (MCBD),

Malvern, Pennsylvania, issued a 10 CFR Part 21 report to the NRC. On November 14, 1986, MCBD issued Service Advice Letter (SAL) 188.1, which instructed recipients to test the relay to identify and correct improper operation.

The SAL stated that MCBD could not detemine when the stop tab problem had begun, but that relays manufactured between January 1983 and October 1986 were suspect.

Date codes were provided that could be 8804110009 111

fN 88-14 April 18, 1988 Page 2 of 4 used to identify potentially affected relays. However, on June 5, 1987, Pilgrim implemented SAL 18'8.1 and identified several improperly operating relays which were manufactured outside the suspected date codes.

In view of this, all HFA relays may be suspect. Additionally, the NRC recently issued NRC Bulletin 88-03, "Inadequate Latch Engagement in HFA Type latching Relays Manufactured by General Electric (gel Company,"

March 10, 1988, regarding problems from inadeouate latch engagement.

b.

PVD PIB and PVD 21D Relays (see Attachment 11 In 1977, GE listed the seismic capability of the normally open contacts of the PVD 21B and PVD 21D differential relays as 6g zero period acceleration (ZPA) with the relay energized and the contacts closed.

The tripping of each of these relays in response to differential over current is by a "low set" voltage unit, device 87L, and a "high set" current unit, device 87H.

Device 87L is an instantaneous voltage unit that has its coil connected across the de terminals of a full wave rectifier device. Device 87L has two sets of nonnally open contacts, one of which is connected in parallel with the contacts of the seal-in unit.

Device 87H is an instantaneous overcurrent Unit, with its coil connected in series with thyrite discs; it has a single set of normally open contacts.

Originally, GE published data which implied that the seismic capability of 6g ZPA applied equally to both 87H and 87L functions.

In 1983, GE revised the seismic capabilities as follows:

the capability of the 87L function remained at 6g ZPA, and the capability of the 87H function was reduced to 2g ZPA. GE stated that the 87H set of normally open contacts is generally wired in parallel with both the open 87L contacts and the open 87L seal-in contacts when the PVD relay is used to trip a circuit brea ke r.

GE stated that the 87H function of the PVD is less sensitive than the 87L function (it requires more operating signali.

As a result, opening the nonnally open 87H contacts for a time greater than 2 milli-seconds as a consequence of a seismic event between 2g and 6g ZPA should not interfere with tripping the circuit breaker because the parallel 87L contact would remain closed up to 6g ZPA.

In view of the ?g ZPA seismic capability of the 87H contact, GE recontrended the application be checked if the 87H centacts are used separately for any functions other than tripping.

c.

HGA Relays (see Attachment 1)

During a re-evaluation of qualification data for relays used in nuclear Class 1E systems, MCPD determined that the seismic data for HGA 11 and HGA 111 relays published in MIL No. RS77-3, dated March 25, 1977, were in error.

The incorrect data were revised in MIL No. 8P-1?, dated July 26, 1982.

Specifically, MIL No. RS77-3 listed the seismic capa-bility of the nonnally closed contacts of ac-rated HGA 11 or HGA 111 112

IN 88-14 April 18,1988 Page 3 of 4 relays at 1.0g ZPA when the relays were in the de-energized state.

The de-rated relays were listed at 1.5g ZPA for the same condition. The re-vised seismic capability of the nomally closed contacts of both ac and dc HGA 11 and HGA 111 relays in the de-energized state is less than 0.5g ZPA.

GE has conducted studies to determine design changes that could improve the seismic capability of the nomally closed contacts when the relay is de-energized. Additionally, a separate study considered the seismic capa-bility of alternate PGA relay models as a means for improving the seismic capability of normally closed contacts with the relay de-energized.

At this time, GE has no plans for development of an HGA relay with improved seismic capability.

On the basis of the above infomation, GE reconnends that Class 1E cir-cuits using nomally closed contacts of HGA 11 and HGA 111 relays, with the relay de-energized, be evaluated for the effect on the entire circuit of an PGA 11 or HGA 111 contact opening longer than 2 milliseconds.

On June 24, 1987, Duane Arnold reported to the NRC (1) that, based on the revised data, PGA 11 relays providing critical functions in the core spray, low pressure coolant injection, high pressure coolant in.iection, and reactor core isolation cooling systems at Duane Arnold have indeteminate qualification, and (2) that the relays were being replaced.

Discussion:

a.

HFA Relays For HFA relays, in addition to SAL 188.1, GE issued the following SALs:

(1)

SAL 139.2, dated April 28, 1976. This SAL superseded SAL 139.1, Jated September 5, 1973.

(2)

SAL 152.1, dated April 28, 1976 (3)

SAL 152.2, dated November 3, 1980.

(4) SAL 152.2A, dated March 12, 1982.

(5) SAL 169.1, dated May ?8, 198?.

b.

PVD 21 Relays For PVD 21 relays, GE issued SAL 174.1 dated April 11, 1983.

This SAL provided clarification of the seismic capability of the 87H and 87L

contacts, 113

IN 88-14 April 18, 1988 Page 4 of 4 Most plants using PVD and HGA style relays are subject to the verification of seismic adequary of mechanical and electrical equipment under Unresolved Safety Issue (VSI) A-46 (as outlined in Generic Letter 87-02). However, PVD and HGA style relays have also been found in newer plants that are not subject to Generic letter 87-02 (e.g., Susquehanna Unit 2 LER 86-024-000).

No specific action or written response is required by this information notice.

If you have any questions about this matter, please contact the technical contact listed below or the Regional Administrator of the appropriate regional office.

+kJf-r-

arles E. Rossi, irect r Division of Operational Events Assessment Office of Nuclear Reactor Pegulation Technical

Contact:

K. P. Naidu, NRR (301) 492-0980 Attachments:

1.

General Electric letter to NRC dated February 24, 1983 i

2.

List of Recently Issued NRC Information Notices 114

l i

l GENER AL $ ELECTRIC MANAetMENT SUSINESS otNtut Etterme cowPANY. ros ontAT VALLtT PARKWAY. uALytm PA i DEPARTMENT February 24, 1983 Hr. Richard Oefoung Ofrector of of fice of Inspection and Enforcement Unitad States Nuclear Regulatory Coenission Washington, D.C. 20555

Dear Mr. DeYoung:

Attached is a copy of a letter relating to PYD218 and PYD 21D relays and a copy of a letter relating to HGAll and HGA111 relays addressed to customers we have knowledge of who say be applying these devices in nuclear Installations.

It is our understanding that this fulfills our obligation to the Nuclear Regulatory Comission.

Yours truly, 7

0.[D L. L. Mankoff Manager-Engineering LLM/gs Attachment cc:

E. J. Fierko Malvern-1203

(

T. R. Hacon Malvern 1205 i

M. J. Nrray Malvern-1206 Y. Thomas NRC l

l T

EMV NE 83-758-000 PDR 115

PYD218 and PYD21D SE!5MIC CAPA81LITY During a re evaluation of qualification data for relays used in nuclear 1E systems, some seismic data for PVD 21 relays published in Mll No. R$77-3, March 25,1977 were found to be in error. Tht incorrect data were subsequently revised in part in MIL No. 82-12, dated July 26, 1982. The correct data for these relays art 6 g ZPA for the 87L function and 2 g IPA for the 87H function.

The nomally open contacts of the PVD218 and PVD210 relays were listed in MIL No. R$77 3 as having a 6 g 2PA capability when these contacts wert closed with the relay it: the operate code. During the period Decetter 19,198D through April 27, 1981, a few Nuclear Certification documents were issued with a revised value of 4 g 2PA. Cert!fications issued after April 27, 1981 through January 17, 1983 listed the seismic capattlity at 2 g IPA. For each of the above cases, the seismic capability was implied to apply equally to the 87H and 87L functions, the two tripping codes of the relay.

The seismic cruability of the 87L function was correct at 6 g IPA. The seismic capability of the 87H function shculd have been listed as 2 g IPA.

The 87H nomally open contact is pentrally applied in parallel with the nomally open 87L contact and nomally open 87L seal-in contact when the PVD relay is used to trip a circuit breaker. The 87H function of the PVD is less sensitive than the 87L function, (requires more optrating signal), therefore opening of the nomally open 87H contact for a time greater than two milliseconds as a consequence of a seismic event between 2 and 6 g IPA, should not interfere with tripping the circuit breaker because the parallel 87L contact would remain closed up to 6 g IPA.

If the 87H contact is used separately for any functions other than tripping, the application should be checked in view of the 2 g IPA seismic capability of the 87H contact.

4 116

I HGAll and HGA111 SE!5MIC CAPABILITY. WORMLLY CLOSED CONTACTS l

During a re-evaluation of qualification data for relays used in nuclear 1E systems, some seismic data for HCA11 and HGA111 relays pblished in MIL No. RS77-3. March 25.1977 were found to be in error. The incorrect data were subsequently revised in h!L No. 32-12, dated July 26, 1982.

The nomally closed contacts of AC rated HGAll or HGA111 riisys were listed in i

MIL No. R$77-3 as having a 1 g IPA seismic capability when the rtlays were in l

the de-energized state. The DC rated relays were listed as having 1.5 g IPA capability for the same condition. A value of 0.5 g IPA should have been listed for the normally closed contacts of both relays in the de-energizgd state.

HGA111 relays were recently re tested for seismic capability and the values obtained for nomally closed contacts with the relay de-energized were less than the 0.5 g ZPA low limit of the test equipoent.

Studies are underway to detemine if there are design changes which could increase the seismic capab111ty of the normally closed contacts when the relay is de-energized. In addition, a separate study will consider the seismic capabilities of alternate models of HGA relays as a means for teproving seismic capability of norually closed contacts with the relay de energized.

Based on the preceding infomation, it is reco: vended that Class 1E circuin using HGAll or HGA111 nomally closed contacts with the relay de-energi:ed De evaluated for effect on the entire circuit of an HGAll or HGA111 contact opening exceecing two milliseconds duration.

I 1

117

Attachtrent 2 IN 88-14 April 18, 1988 Page 1 of 1 LIST OF RECENTLY ISSUED NRC INFORMATION NOTICES Information Date of Notice No.

Subject issuance Issued to 88-13 Water Haniner and Possible 4/18/88 All holders of OLs Piping Damage Caused by or cps for nuclear Misapplication of Kerotest power reactors.

Packless Metal Diaphragm Globe Valves 88-12 Overgreasing of Electric 4/12/88 All holders of OLs Motor Bearings or cps for nuclear power reactors.

88-11 Potential loss of Motor 4/7/88 All holders of OLs Control Center and/or or cps for nuclear Switchboard Function Due power reactors, to Faulty Tie Bolts 88-10 Materials Licensees:

Lack 3/28/88 All NRC licensees of Management Controls Over authorized to use Licensed Programs byproduct material.

87-44, Thimble Tube Thinning in 3/28/88 All holders of OLs Supp. 1 Westinghouse Reactors or cps for nuclear power reactors that employ a Westinghouse NS$$.

88-09 Reduced Reliability of 3/18/88 All holders of OLs Steam-Driven Auxiliary or cps for nuclear Feedwater Pumps Caused power reactors, by Instability of Woodward PG-PL Governors 88-08 Chemical Reactions with 3/14/88 All NRC licensees Radioactive Waste generating or pro-Solidification Agents cessing low level radioactive waste.

OL = Operatina License CP = Construction Permit 118

UNITED STATES NUCLEAR REGULATORY C0 MIS $10N OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C.

205 %

April 26, 1988 NRC INFOPMATION NOTICE NO. 88-19:

QUESTIONABLE CERTIFICATION OF CLASS 1E COMPONENTS Addressees:

All holders of operating licenses or construction pemits for nuclear power reactors.

Purpose:

This information notice is being provided to alert addressees to a possible problem with the certification of Class IE components furnished by Planned Maintenance Systems (PMS) of Mt. Vernon, Illinois.

It is expected that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in this information notice do not constitute NRC requirements; therefore, no specific action or written response is required.

Description of Circumstances:

On April 1,1988, Wolf Creek Nuclear Operating Corporation (WCNOC) submitted a written 10 CFR Part 21 Notification to NRC Region IV cencerning 60 Class IE fuses that had been procured from PMS. One of the requirements o' the purchase order (PO) issued to PMS was that the PO items were to be supplied in accord-ance with the requirements of a specific fuse qualification specification for Class IE equipment. This specification contained detailed requirements including materials, environmental qualification, seismic qualification, and inspection /

test requirements. The PPS Certificate of Compliance supplied with the order certified that all of the procurement document requirements had been met and no deviations from the recuirements had been identified.

The fuses were placed on hold by WCN0r. because a required Quality Department surveillar.ce had not been performed. A subsequent WCNOC succe111ance revealed that the records in PMS's possession did not support the statement on the PMS Certificate of Compliance that all PO requirements had been met. Quilification specification requirements were not envered by PMS quality assurance records with respect to infomation on environnental cualification, radiation levels, and seismic qualification.

In addition, it could not be established that a required continuity / resistance check of each fuse had been performed before the fuses were shipped.

08M220217 119

TN 88-19 April 26,1988 Page 7 of ?

Discussion:

The WCNOC 10 CFR Part 71 notification has brought into question the validity of the Certificate of Compliance issued by PMS for Class IE fuses that they supplied. Accordingly, licensees may wish to review Class IF component pro-curements from this vendor to ensure that appropriate bases exist for the use of the components.

No specific action or written response is required by this infonnation notice.

If you have any questions about this matter, please contact the technical contact listed below or the Regional Administrator of the appropriate regional office.

L$.

Charles E. Rossi, Director l

Division of Operational Events Assessment Office of Nuclear Reactor Regulation Technical

Contact:

Joseph 1. Petrosino, NRR (301) 492-0979

Attachment:

List of Recently Issued NRC Information Notices i

120

Attachment IN 88-19 April 26,1988 Page 1 of 1 LIST OF RECENTLY ISSUED l

NPC INFORMATION NOTICES l

Information Date of I

Notice No.

Subject Issuance Issued to 88-18 Malfunction of Lockbox on 4/25/88 All NRC licensees Radiography Device authorized to manufacture, distribute, and/or operate radio-graphic exposure devices.

88-17 Summary of Responses to NRC 4/22/88 All holders of OLsBulletin 87-01, "Thinning of or cps for nuclear Pire Valls in Nuclear Power power reactors.

Plants" 88-16 Identifying Waste Generators 4/22/88 Radioactive weste in Shipments of Low-level collection and Waste to Land Disposal service company Facilities licensees bandling i

prepackaged waste, and licensees operating low-level waste disposal facilities.

88-15 Availability of U.S. Food 4/18/88 Medical, Academic, and Drug Administration and Conmercial (FDA)-Approved Potassium licensees who possess Iodide for Use in Emergencies radioactive iodine.

Involving Radioactive Iodine 88-14 Potential Problems with 4/18/88 All holders of OLs Electrical Relays or cps for ruclear power reactors.

1 88-13 Water Hanner and Possible 4/18/88 All holders of Ols Piping Damage Caused by or cps for nuclear Misapplication of Kerotest power reactors.

Packless Metal Diaphragm Globe Valves 88-12 Overgreasing of Electric 4/12/88 All holders of Ols Motor Pearings or cps for nuclear power reactors.

OL = Operating License CP = Construction Permit 121

INDEX FACILITY REPORT NUMBER PAGE Allegheny Ludlum Corp.

Claremore, Oklahoma 99902007/88-01 1

American Insulated Wire Corp.

Pawtucket, Rhode Island 99900399/88-01 7

Circleville Metal Works Circleville, Ohio 99901075/88-01 13 General Electric Company Wilmington, North Carolina 99900003/87-01 21 GE Nuclear Energy San lose, California 99900403/87-06 27 Nutherm International Inc.

Mount Vernon, Illinois 99900779/87-01 57 Pullman Construction Industries, Inc.

Chicago, Illinois 99900853/87-01 81 l

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SIBUOGRAPHIC DATA SHEET NUREG-0040 Vol. 12, No. 1 us it. vet.o.s o.,

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..w gy Licensee C tractor and Vendor Inspection Statu s g'

Rsport, Qua terly Report - January 1988 thru 8

March 1988

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1988 i.wv-oa

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Division of Reacto Inspection and Safeguards

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1 Office of Nuclear

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Nuclear Regula,ry Commission

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Washington, D.C.

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This periodical covers the re Itg f inspections performed by the NRC's Vondor Inspection Branch that h e been distributed to the inspected org anizations during the peri

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Also included in this issue a the results of certain inspections performed prior to January 1 8 $ bat were not included in previous issues of NUREG-0040.

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UNITED STATES

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NUCLEAR REGULAT8RY C8MMISSION Posvace,*yis Po'o WASHINGTON, D.C. 20555 OFFICIAL BUSINESS PEN ALTY FOR PRIVATE USE,6300 120555078877 1 1ANINV US NRC-0 ARM-ADM DIV 0F PUB SVCS POLICY & PUB MGT BR-PDR NUREG W-537 WASHINGTON OC 20555 1

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