IR 05000313/1972012
| ML19317H093 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 01/24/1973 |
| From: | Brownlee V, Crossman W, Long F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19317H077 | List: |
| References | |
| 50-313-72-12, 50-368-72-08, 50-368-72-8, NUDOCS 8004140576 | |
| Download: ML19317H093 (17) | |
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. - ' ~ O , . ) p,t'0'co A rs UNITED STATES /V ' y h'- -)9 . ATOMIC ENERGY COMMISSION t
- J' DIRECTORATE OF REGULATCRY CPERATICUS ' _( REGION 11 - SusT E 818 2M A E AC MT REE ST R EET. NORT MwEST tc.co.c.s.
ac.*saeases AT I A NT A. GEORGI A 30303 RO Inspection Report Nos. 50-313/72-12 and 50-368/72-8 Licensee: Arkansas Power and Light Company Sixth and Pine Streets Pine Bluff, Arkansas 71601 Facility Name: Arkansas Nuclear One, Units 1 and 2 Docket Nos.: 50-313 and 50-368 License Nos.: CPPR-57 and Pending Category: A3/B1, Al Location: Russellville, Arkansas Type of Licensee: B&W, PWR-2568 ht, 880 Mwe Combustion, PWR-2910 ht, 990 Mwe Type of Inspection: Announced Dates of Inspection: November 28 - December 1, 1972 Dates of Previous Inspection: October 25-27, 1972 Principal Inspector: V. L. Brownlee, Reactor Inspector Facilities Section Facilities Construction Branch Accompanying Inspectors: W. A. Crossman, Acting Senior Inspector Facilities Section Facilities Construction Branch.
Other Accompanying Personnel: None /13 35 . Principal Inspector: )/bi ' ~ ' ' ' - , V. L. Brownlee, Reactor Inspector f Dat#- Facilities Section Facilities Construction Branch.
Reviewed By E QM__ l7d!"/3 F..f.
Long, Chilf, Facilities Construction lDMe Branch i s t l
8 0 0 41405 7g l .
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' s RO Report Nos. 50-313/72-12-2- - , and 50-368/72-8 i SUMMARY OF FINDINGS - I.
Enforcement Action A.
Violations ! None . B.
Safety Items None II.
Licensee Action On Previously Identified Enforcement Matters A.
Violations None B.
Safety Items N , None III.
New Unresolved Items , At the conclusion of this inspection, two items remained unresolved ' and require further examination by the applicant and Region II.
! 72-12/1 Discrepancies Between Piping and Control Systems (Unit 1) ! It is our understanding that AP&L will provide an engineering evaluation regarding the extent of the control system cross connec-tion.
AP&L will be required to document the deficiency, analysis of safety implications and corrective action if safety related systems are identified which have control system cross connections.
(Details I, paragraph 7) 72-8/l Reactor Coolant Pump Guide Vanes (Unit 2) It is our understanding that JP&L will provide additional informa-tion regarding ANO-2 primary coolant pump guide vane installation.
(Details I, paragraph 4) IV.
Status of Previously Reported Unresolved Items The following list of unresolved items represents a summary of all outstanding unresolved items identified to the applicant by inspec- \\ tion report letters or letters from the Regional Director, Region II.
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_ . . . - _ - _ _ - _ _ _ _ . . - - .RO, Report Nos. 50-313 72-12-3- / - and 50-368/72-8 - 71-6/1 Stressing Washee Design (Unit 1) (Region II Letter, January 24, 1972, Item 1) Undesirable bending of the tendon wires was observed whilu the stressing washer was being advanced on the tendon assembly for buttonheading.
Since this may relate to the design of the stressing washers, the items has been referred to the Directorate of Regulatory Operations, Head-quarters for further evaluation. The licensee will be informed of the results of this evaluation when completed.
This item remains open.
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' 72-2/1 In-Core Instrumen Guide Tube Failure (Unit 1) (Region II Letter, May 23, 1972, Item 1) It is our understanding that the licensee will determine the applicability to his facility and conduct an evalua-tion of the failure of reactor in-core instrument guide , tubes which occurred at a similar facility. This item remains ' open.
72-4/1 Reactor Vessel and Reactor Vessel Internals Modification s_, (Unit 1) (Region II Letter, July 27, 1972, Item 1) It is our understanding that a design change is in prepara-tion regarding this item. We have referred the item to Regulatory Operations, Headquarters for further evaluation and will advise the licensee the results of this evaluation at a later date. This item remains open.
72-4/2 Cable Routing (Unit 1) (Region II Letter, July 27, 1972, Item 2) The routing of cables under the control room removable floor e.ty not meet safeguards and reactor protection systems seftrations and loading criterio.
It is our understanding that the licensee will evaluate this item and corrective , measures will be initiated as may become necessary.
' This item remains open.
72-4/3 Cable Routing (Unit 1) (Region II Letter, July 27, 1972, Item 2) The separation of safety channel wiring inside the Babcock and Wilcox Company (E&W1 supplied panels may not meet the /'~ requirements of the safeguards and reactor prote.ction systems-( separations and loading requirements.
It is our understand-ing that the licensee will evaluate this item and corrective l l ._.
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. . _ - 'D(h RO Report Nos. 50-313/72-12-4-and 50-368/72-8 measures will be initiated as may become necessary.
This item remains open.
72-4/4 Control Rod Drive Motor Tube Defect Analysis (Unit 1) (Region II Letter, July 27, 1972, Item 3) The control rod drive motor tube housings contained metal defect indications and possible thin walls.
It is our understandine that design changes are being made and will be the subjer.t of a B&W topical report. This item has been referred to Regulatory Operations, Headquarters for further evaluation. We will advise the licensee of the results of this evaluation when it becomes available.
This item remains open.
72-7/1 Pressurizer Safety and Relief Valve Mounting and Connecting Piping (Unit 1) (Region II Letter, August 21, 1972, Item 1) It is our understanding that the licensee will evaluate the s pressurizer safety and relief valve mountings and connecting ) piping to assure that these have been analyzed for dynamic gN- / reaction forces during blowdown. This item remains open.
72-8/l Containment Structure (Unit 1) (Region II Letter, October 26, 1972, Item 1) It is our understanding that the licensee will continue the evaluation of the possibility of damage te the containment structure as a result of the stressing jack failures during stressing operations. This item is closed.
(Details I, paragraph 5) 72-11/1 Paddle-Type Flow Switches sJnit 1) (Region II Lettert December 19, 1972, Item 1) It is our understanding that the licensee will obtain infor- , mation regarding the extent to which paddle-type flow switches are utilized within the nuclear piping systems.
This item remains open.
72-11/2 B&W Safety Cabinets - Internal Panel Wiring (Unit 1) (Region II Letter, December 19, 1972, Item 2) (Sea Item 72-4/3.)
It is our understanding that the licensee will require-s i that B&W perform an engineering evaluation of all safety ,_s/ cabinets regarding conformance to the separations criteria for internal cabinet wiring. This item remains open~. . - - -
. _ _ _. _ _ _ .. -. _ - ... _ . . _ i . .RO Report Nos. 50-313/72-12-5- ,and 50-368/72-8 . , 72-11/3 Cable Installation in Control Room and Computer Room False
Floor and Floor of Main Control Panels (Unit 1) (Region II ! Letter, December 19, 1972, Item 3) (See Item 72-4/2.)
It is our understanding that t.he licensee will require Bechtel Engineering Corporation (Bechtel) to evaluate cable installation under the control room and computer room false floor regarding separations, loading and quality ' workmanship criteria. This item remains open.
72-11/4 Aluminum Instrumentation Housings Located Inside Containment (Unit 1) (Region II Letter, December 19, 1972, Item 4) It is our understanding that the licensee will determine that the aluminum instrument cases within containment are accounted for in the containment aluminum inventory.
This item remains open.
(Details I, paragraph 6.)
' 72-11/5 Storage and Inspection Program For Installed or Stored ' In-Place Equipment (Unit 1), (Region II Letter, December 19, 1972, Item 5) , It is our understanding that the licensee will determine that Bechtel does have and is implementing a quality assurance storage and inspection program for installed or stored in-place equipment. This item remains open.
72-11/6 General Electric Company (G-E) Reactor Trip Breakers , (llnit 1) (Region II Letter, December 19, 1972, Item 6) ! It is our understanding that the licensee will obtain test data for our review on the G-E reactor trip breakers (four). This item remains open.
72-11/7 Qualification of Containment Instrumentation (Unit 1) (Region II Letter, December 19, 1972, Item 7' It is our understanding that the licensee will confirm that instruments located within the containment building were designed and qualified to function following the Design Basis Accident (DBA). This item remains opan.
The following listed items are included in this report as New Unresolved Items for purposes of identification only. These items were previously identified to the licensee by letter from the Region II Regional Director as indicated: ! l --- . - - - -. -... - - . ... -
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. % R0 Report Nos. 50-313/72-12-6-and 50-368/72-8 72-8/2 _ Reactor Scram Circuit Breakers (Unit 2) (Region II Letter, December 9, 1971) It is our understanding that AP&L will determine whether
- 5= scram breakers to be installed are of the make and model deceribed in the RO:II letter.
If the answer is affirmative, AP&L will provide the date when each reactor scram breaker will be inspected and tested. This item re-maine open.
(Details I, paragraph 9) 72-8/3 Steam Line Pressure Relief Valve Headers (Unit 2) (Region II Letter, December 10, 1971) It is our understanding that AP&L will provide a main steam line pressure relief valve manifold stress analy-sis and engineering evaluation. This item remains open.
(Details I, paragraph 9) 72-12/2 Valve Wall Thickness Verification (U'it 1) n (Region II Letter, June 30, 1972) It is our understanding that AP&L will develop and implement a verification program when AEC defines the requirements regarding their letter of exception, July 28, 1972. This item remains open.
(Details I, paragraph 9) 72-8/4 Valve Wall Thickness verification (Unit 2) (Region II Letter, June 30, 1972) ' It is our understanding that AP&L will develop and imple-ment a verification program when AEC defines the require-ments regarding their letters of exception, July 28 and August 7, 1972. This item remains open.
(Details I, paragraph 9.)
V.
Design Changes None .A.
Unusual Occurrences None VII.
Other Significant Findings 'N ) A.
Project Status ss_/ - l See Details I, paragraph 2.
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' . - . . f .R0 Report Nos. 50-313/72-12-7- '50-368/72-8 i B.
Personnel or Organization Changes No significant organizational or functional alignment changes.
C.
Inspection Effort ! This inspection was performed in conjunction with an inspec-tion of the electrical and control systems installation by Licensing Representatives Bernero, Calvo and Ippolito.
VIII.
Management Interview The inspectors met with AP&L and constructor management personnel and apprised the attendees of the general areas of inspection and findings.
Two new unresolved items were identified (see Section III, ' New Unresolved Items) as requiring further examination during subsequent inspections.
j l The licensee was informed that Regulatory Operations had no l further questions at this time regarding the previously reported unresolved item 72-8/1, Containment Structures.
i Examination of records of fabrication and erection of the critical l piping systems was discussed with the licensee.
The depth and scope of the inspection in this area were outlined.
' The licensee was informed that the discrepancies observed in the shop records were resolved prior the management meeting.
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_m._ _ _ ._ . _. . ' . _ ,RO Report Nos. 50-313/72-12 I-l and 50-368/72-8 DETAILS I Prepared By: '/[ ,'
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. V. L. Brownlee, Date Reactor Inspector, Facilities Section, Facilities Construction Branch [[d / ZI
- W. A. Crossman, 'Date Acting Senior Inspector, Facilities Section, Facilities Construction Branch Dates of Inspection: November 28 - December 1, 1972 Reviewed By k ~ ' %_/ I z.4!73 F. J. Long, Chief Date' Facilities Construction Branch All information in the details applies equally to ANO-1 and 2 except l where information is identified with a specific reactor.
1.
Individuals Contceted a.
Arkansas Power and Light Company (AP&L) N. A. Moore - Chief QA Coordinator A. C. Bland - QC Inspector (Civil) C. L. Bean - QC Inspector ()fechanical) E. Quattlebaum - QC Inspector (Electrical) G. H. Miller - Assistant Superintendent W. Cavanaugh - Project Manager J. Grisham - Assistant Engineer , J. McAlister - Production Engineer l D. Reuter - Assistant Engineer b.
Bechtel Engineering Corporation (Bechtel) W. T. Stubblefield - Project Superintendent J. B. Lotit - Project QA Engineer C. G. Beckham - QA Engineer (Mechanicall R. E. Allen - QA Engineer (Electrical) l O P. W. Sly - QC Engineer (Recordsl ! Q K. Higgens - QC Engineer ()fechanical) i - . - --. - - - - .- . -
_ - _ _ - _ - - _ _ _ - - . O R0 Report Nos. 50-313/72-12 I-2 and 50-368/72-8 G. Katanics - Project Engineer J. Haidinger - Engineer, Electrical Group Supervisor W. Kunz - Senior Engineer, Electrical F. Silberman - Lead Engineer, Instrumentation J. Oszewou - Licensing Engineer B. F. Allen - Senior Field Welding Engineer W. D. Schuster - QC Field Welding Engineer c.
Babcock and Wilcox Company (B&W1 E. Baker - Associate Project Manager A. B. Lloyd - Systems Engineer C. A. Stemke - Systems Engineer 2.
General Progress of Construction: The following list provides the estimated percent physical construction complete total effort: \\ Description 10/27/72 12/1/72 Unit No.1 (Total)
91 Piping Systems
94 Reactor Coolant
97 Core Flood
80 Decay Heat Removal
96 Makeup and Purification
85
90 Main Steam - Feedwater
93 Electrical (Total)
88 Raceway and Conduit
98 Switchboard and Shutdown Board Installation
98 , Cable Pull
90 ' Cable Termination
80 Reactor Protection
75 Engineered Safety Features
8Q Switchyard
91 Unit No. 2 (Total)
o Total Effort Under Exemption
97 O>g , . .- --
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. _ ! '. + .. RO Report Nos. 50-313/72-12 I-3 and 50-368/72-8 Bechtel personnel onsite for Unit No. 1 is 792 and 53 subcontractor personnel; Unit No. 2, 84 and 93 subcontractor personnel; total, 1022. AP&L personnel 92 and 5 QA personnel. Labor problems, none.
3.
Work Summary and Schedule . a.
Unit 1 i Site efforts continue to be concentrated on piping and electrical
systems installation. All major components have been set.
Electrical systems installation (cable pulling and termination) is progressing at full capacity.
Systems turnover remains slow.
Systems completely turned over, i ! 50%; partially turned over, 32%; remaining fully in construction I status, 20%. Some preoperational testing is reported to have started.
l ! b.
Unit 2 ' I \\~- Exemption work is virtually complete.
Some work efforts r.ontinue in the turbine building and cooling tower area.
! ! 4.
Reactor Coolant Pump, Guide Vane Failure (Unit 2, Inquiry Report l
No. 50-309/72-01) ! ! AP&L'has generated a letter to the Combustion Engineering, Inc. (CE) project manager requesting that CE provide AP&L with. a description I of the Maine Yankee incident, an analysis of safety implications, a comparison of the ANO-2 and Maine Yankee pumps, and what corrective action will be taken for ANO-2 components.
Followup of this item will be included in the scope of future inspections.
i ! 5.
Concrete Testing in the Area of Tendon 2?.-H.-32 Unit 1 Containment, Elevation 437 l Concrete was tested by Letcher and Associates. Their findings indicate that no major flaw areas exist in the area being tested.
' Tests were performed on buttress No. 1 at elevation 467 and on j buttress No. 1 at elevation 437. Additional tests were performed at buttress No. 2 at elevation 437. A visual examination of the ' tendon area between buttress No. 1 and 2 showed no apparent cracks.
The Bechtel supervising materials engineer, San Francisco, witnessed
the testing and concludes that the concrete is of good sound quality and no further testing is required.
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_ _ _ __ _ _ _ _ _ _.. _ _ _ _ _ _ ____ _ _ d . . -
. RO Report Nos. 50-313/72-12 I-4 a ' and 50-368/72-8 i t
Examination of site records and discussion with site QA/QC personnel indicate that Regulatory Operations has no further questions at this time.
i 6.
Aluminum Instrumentation Housings on Instruments Located Inside Containment (Unit 1, RO letter - December 19, 1972. Item 72-11/4) ' Section 6.6, pages 6-16e and f of the PSAR, specifies that instrument transmitter covers located within the reactor building be painted with a qualified coating that will not react following the DBA.
Bachtel has requested that B&W provide documentation showing housing i ma arial and protective coating used for Bailey, Foxboro and Motorola tram mitters.
i 7.
Discrepancies Between Secondary Loop "A" and Loop "B" Piping and , l Control Systems , Numerous cross connections have been identified between secondary j loop "A" and "B" piping and control systems. The problem occurred j when steam generator "A" was placed in primary coolant loop "3" and steam generator "B" was placed in primary coolant loop "A" per
engineering drawing. To accommodate this change, all associated , I ancillary piping, equipment, and instrumentation in loop "A" (such as feedwater pump) received a "B" suffix to be consistent with the steam generator suffix. The changing of component designation and lack of fo:lowup of engineering evaluation and document chanFes resulted in instruments located on the steam generator and piping
associated with the A primary loop (B steam generator) are perform-ing control functions on the south secondary (A steam generator) loop.
AP&L/Bechtel reported that their preliminary evaluation sa not l been concluded at this time and they were not prepared to provide a statement as to whether or not this crossoeer proble:a may enend to safety related systems. Regulatory Operations vas assured that
the crossover problem was being evaluated concerning its safety j implications.
l-l This item is not considered at this time, by AP&L, to be reportable
to the AEC as a significant deficiency as required by 10 CFR 50.55(.e).. AP&L assured Regulatory Operations that significant findings associated with safety related systems identified during the evaluation wou'd i be evaluated as an AEC reportable item, ! Bechtel records had identified this as a significant deficiency.
No onsite rework has started.
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- - - . _ ._ __- _- _ _ _ _ _ _ _ _ _ . . . . i . RO Report Nos. 50-313/72-12 I-5 and 50-368/72-8 i , i ' . ' Regulatory Operations will include followup of this item within the scope of future inspections.
8.
Resctor Vessel and Internals Modification < Modifications to the reactor vessel and internals is scheduled for completion, December S, 1972.
9.
New Unresolved Items The following listed items are made a part of this report to identify items that remain unresolved that were originally identified to the licensee by letter from the Region II Regional Director and not a result of inspection report letters.
a.
Reactor Scram Circuit Breakers (Unit 2) (Region II Letter, December 9, 1971) This item is given identification No. 72-8/2.
(See Summary of Findings, Section IV of this report.)
b.
Steam Line Pressure Relief Valve Headers (Unit 2) (Region II Letter, December 10, 1971) Ihis item is given identification No. 72-8/3.
(See Summary ' of Findings, Section IV of this report.)
I c.
Valve Wall Thickness Verification (Unit 1) (Region II Letter, June 30, 1972) This item is given identification No. 72-12/2.
(See Summary
of Findings, Section IV of this report.)
d.
Valve Wall ThicAness Verification (Unit 2) (Region II Letter, June 30, 1972) This item is given identification No. 72-8/4.
(See Sunrnary of Findings, Section IV of this report.} 10. Piping Systems Erection Eleven joints randomly selected from four critical piping systems were examined as a followup effort to verify compliance with the field QA/QC program for erection of piping systems.
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. . N ) ,R0 Report Nos. 50-313/72-12 I-6 and 50-368/72-8 . These joints as related to the systems are as follows: Main Steam Weld No.
Kellogg Iso.
Pipe Size Wall Thickness EBB-3-2A 1-MS-1 36" 1.055" 3-3 1-MS-1 26" .762" 3-5 1-MS-1 26" .762" 3-16 1-MS-3 36" 1.055" 3-17 1-MS-3 36" 1.055" Core Flood Weld No.
Kellogg Iso.
Pipe Size Wall Thickness CCA-6-14 6-CF-2 14" Schedule 10 6-5 6-CF-1 14" Schedule 10 Decay Heat Weld No.
Kellogg Iso.
Pipa Siza 3.1 Thickness-CCB-1-13 7-DH-2 12" Schedula 10 CCB-1-6 7-DH-1 12" Schedula 10 Make Up Weld No.
Kellogg Iso.
Pipe Siza Wall Thickness CCA-5-5 17-MU-23 2" Schedula 160 5-2 17-MU-23 2" Schedula 160 Comments relative to a specific joint within the areas examined are in regard to exceptions only, a.
Records of QC Inspection Records of QC inspector verification of visual inspection, heat treatment, NDT, welding material control and weldor identity appear on the weld traveler. Examination of these documents revnaled proper identification inspection points by the QC welt!ing engineer. Hold points for code compliance by the insurance inspector were designated at random and properly signed off. Final signoff by the senior velding engineer , l 'N was noted.
' No significant deficiencies were identified.
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. - y/ . RO Report Nos. 50-313/72-12 I-7 and 50-368/72-3 b.
Heat Treatment Records Preheat and stress relief requirements ware designated and satis-factory completion certified by signoff of the weld travelers and PWHT strip charts for the main steam system welds.
Heatup rate, soak time, and cooldown rate were identified on the temperature strip charts.
No defic 1encies were identified.
c.
Nondestructive Testing Records Film review was made for both shop and field welds.
The radiographs examined for Kellogg shop welds were for sub-assembly of four spool pieces from the decay hert, makeup, core flood and main steam piping systems.
Austenitic and carbon steel materials were involved.
\\ Radiographic interpretation records for all final welds including weld repairs were observed to properly identify the weld and correlate to procedure, radiographer, reviewer, acceptor, etc. Radiographs were of quality required by code. Proper identi-fication of film and weld were observed. Penetrameter utili-zation and placement were in accordance with code. No significant discrepancies were observed.
d.
Qualification of Weld Procedures Weld procedures indicated on the field welding checklists were the latest revisions available to the field at the time of welding.
Qualification of the procedures have been reviewed and previous deficiencies resolved.
e.
Qualification of Weldors Performanca qualification records for the designated weldor were examined and revealed no deficiencies.
It was observed that tha performance for the weldor's qualifi-cations wera current, f, Pftysical Appearance of Completed Weld l T!ie selected welds were examined visually for physical appear-l
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. - ., ) .RO Report Nos. 50-313/72-12 I-8 . ' ' and 50-368/72-8
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ance of the completed weld on the 0.D.
No cracks, undercut or surface porosity was observed. Weld profile examination revealed good blending to base material and proper weld crown ! for the wall thickness.
11. Shop Fabricated Piping Materials i Spool pieces were randomly selected from two systems for examina-tion to verify compliance witit the shop QA/QC program for fabri-cation and testing of subassemblies.
Joints in regard to their respective systems are as follows: ' Decay Heat (System #7) i Kellogg Spool Piece Pipe Size ASTM Wall Thickness I DE-HCB-A4-74 14" A 312 Schedule 10 (ice 11ogg 150 ELL A 240TP304/A403 + ( 7-DH-13) ( (Shop Fab #F-199 l l Decay Heat (System #7) Kellogg Spool Piece Pipe Size ASTM Wall Thickness , DE-HCB-A6-69 10" A 358 CL1 Schedule 10 (Iso 7-DH-12) 14" A 312 (Shop Fab No. F205 Make Up (System #17) Kellogg Spool Piece Pipe Size ASTM Wall Thickness MU-CCA-A4-71 2h" Type 316 Schedule 160 (Iso 17-MU-23) 312 SMLS (.375) (Shop Fab. No. F-464 TO B31.7 i Main Steam System (System #1) i Kellogg Spool Piece Pipe Size ASTM Wall Thickness
MS-EBB-AS-11 26" A155, .793 (Kellogg Iso 1-MS-3) 36" Gr. KCF70 1.055" (Shop Fab. No. F-63) CL 1
d
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. A ,RO Report Nos. 50-313/72-12 I-9
and 50-368/72-8 a.
Material Certification Records Data packages from Kellogg for eacit spool piece assembly contain shop assembly sheets whicit cross-reference the assembly mark number and Kellogg isometric drawings.
Material test reports are traceable from the assembly drawings by " Material Report Code Symbol." Components comprising the assembly are identified on the test report by heat number. Chemical composition and physical characteristics for the components which comprised the assemblies were examined.
It was observed that no ferrite content was noted for Kellogg inserts, KI-7.
The Shop Assembly Sheet specifies 5 to 25% ferrite. This item was resolved by location of a missing sheet accompanying the mill certs.
For spool piece DH-HCB-AG-69, the QA statement specifies a welding procedure for carbon steel for welding a forged stain-less coupling to the stainless steel assembly, specifically, the joint designated as "K" and the procedure used was identi-fied PI-SV-F4-SMAW-4-5G.
This discrepancy was resolved as a " typo" error.
5.
NDT and Examination During Fabrication
Records identifying nondestructive testing and inspection during fabrication are in the form of Assembly Fabrication , Sheets ("F" Sheets) and " Quality Assurance Statements." These i k documents describe required testing and certification of satis-factory results by both Bechtel shop inspector and the Kellogg senior QC engineer.
Proper signoff was observed in the cases reviewed. No { deficiencies were observed.
c.
Hydrostatic Testing Hydrostatic test results for individual components and piping , is certified by material suppliers in accordance with the specified ASTM designations.
No discrepancies were noted.
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