IR 05000364/1993400

Insp Rept 50-364/93-400 on 931004-15.No Violations Noted. Majors Areas Inspected:Review of Licensee ISI Program Submittal & Nondestructive Exams of safety-related Weldments,Hanger & Supports Selected from SI

ML20059B730
Person / Time
Site:	Farley
Issue date:	11/10/1993
From:	Gray E, Harris R, Modes M, Peterson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20059B707	List: IR 05000364/1993400 ML20059B704
References
50-364-93-400, NUDOCS 9401040230
Download: ML20059B730 (11)

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Enclosure

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

DOCKET / REPORT NO /93-400 LICENSE N NPF-8 l

LICENSEE: Alabama Power Company

' 600 North 18th Street Birmingham, AL 35203 FACILITY NAME: Joseph M. Farley Nuclear Plant l

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INSPECTION AT: Ashford, Alabama INSPECTION CONDUCTED: October 4,1993 through October 15, 1993

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CONTRACTORS: William Mingus, TET,Inc., Mobile, AL

! David Payne Jr., TET, Inc., Mobile AL INSPECTORS: [ M //// 3

/ Dat'e

'Rpard ILIfar(s, DE Technician

< Mobile NDE Laboratory, EB, DRS

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& ///8/f3 Patrick M. Peterson, NDE Technician 'Date l Mobile NDE l2boratory, EB, DRS APPROVED: // 73 E. Harold Gray, Chief Date Mobile NDE 12boratory, EB, DRS

// > lf T Kiichael C. Modes, Chief ' IIate '

Materials Section, EB, DRS I

9401040230 931223 PDR ADOCK 05000348 *

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Inspection Summary: An announced inspection was conducted at Joseph M. Farley Nuclear Plant (Farley) during the period October 4,1993, through October 15, 1993, using the NRC's Mobile Nondestructive Examination (NDE) Laboratory, (Report No. 50-364/93-400).  ;

The purpose of the NDE Mobile Laboratory is to perform independent nondestructive examinations, evaluations of components, systems and weldments to assure that examinations performed are in compliance with codes, standards and regulatory requirement '

Areas Insnected: Areas examined during this inspection included a review of the licensee inservice inspection (ISI) program submittal and nondestructive examinations of safety-related weldments, hanger and supports selected from the safety injection (SI), residual heat removal (RHR), pressurizer (PZR), steam generator "C," and feedwater (FW) systems. Also included in this inspection, was a review of the erosion / corrosion (E/C) progra ,

Enults: Within the expected normal variations in examination techniques, the result of the NDE evaluations performed by the NRC essentially agreed with the results obtained by the :

licensee with one exception. The NRC did not agree with the licensee's disposition of an indication in the feedwater pipe to nozzle weld APR2-4350-22. The NRC concluded, from the results obtained by its independent evaluation, that the indication was a thermal fatigue crack. The licensee disposition of this indication is that it is undercut at the weld root but stated the area would be reexamined by ultrasonic examination (UT) at the next refuel outage, or sooner, if an outage prior to then presents an opportunity for reexaminatio The inservice inspections, evaluated by the NRC, were thorough and in full compliance with '

the requirements of the Federal Code and ASME Section XI. The program for inspection is staffed by qualified, professional personnel. The individual inspections performed were i conservatively execute l

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DETAILS INDEPENDENT MEASUREMENTS - NRC NONDESTRUCTIVE EXAMINATION AND QUALITY RECORDS REVIEW OF SAFETY-RELATED SYSTEMS (73753)

During the period of October 4, through October 15,1993, an onsite independent inspection was conducted at Farley Nuclear Plant. The inspection was conducted by NRC inspectors and contractors. The objectives of this inspection were to assess the adequacy of the licensee's inservice inspection program, the licensee's actions regarding the as-built configuration of pipe hanger / supports, and implementation of the erosion / corrosion progra This was accomplished by duplicating those examinatious performed by the licensee, required by regulations and codes, evaluating the results, and performing a detailed review of the ISI program, E/C program and NDE procedures used to implement these programs. Section of this report contains a listing of the specific weld, hangers and supports inspecte The Code of Federal Regulations (CFR), Title 10, Part 50.55a (10 CFR 50.55a), requires j inservice inspection (ISI) of safety-related equipment to identify system degradation. Before the licensee generated program of inspection is applied to the equipment, it must be approved by the Nuclear Regulatory Commission (NRC) under the authority embodied in 10 CFR 50.55a (g) (4) (iv). The required inspections are detailed in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, for Inservice Inspection as embraced in 10 CFR 50.55a (b). The NRC inspection described in this report was made using the Mobile Nondestructive Examination (NDE) I2boratory. The Mobile NDE 12boratory is capable of independently performing the examinations required of the licensee. This capability provides the NRC with unique insights into the licensee's inservice inspection program and on a sampled basis, the adequacy and accuracy of the licensee's .;

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specific NDE inspection The scope of this inspection was to review the administrative portions of the program and to perform NDE of the systems that were availabl .1 Nondestructive Examination (NDE)

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Inspection Hanger / Support (57050)

Eleven (11) safety-related hanger / supports were visually inspected per NRC Procedure NDE-10, Rev. O, Appendix A and B, in conjunction with Farley Nuclear Plant Site Procedure FNP-0-NDE-157.3, Rev. 2, and quality control documents and associated '

isometric / drawings. Included in this inspection were hanger / supports selected from the SI

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system. The accessible surface area and adjacent base metal for a distance of one-half inch on either side of the weld was examined. Component integrity was also examined, including proper installation, configuration or modification of supports, evidence of mechanical or structural damage, corrosion, bent, missing or broken member Results The inspections by the NRC closely matched those of the licensee. No deviations j were identifie '

Visual Examination (VT) (57050)

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Twenty (20) safety-related pipe weldments and adjacent base material (1/2 inch on either side i of the weld) were visually examined in accordance with NRC Procedure NDE-10, Rev. O, ;

Appendix A, Site Procedure FNP-0-NDE-157.3, Rev. 2, quality control documents,  !

isometrics and as-built drawings. Examined during this inspection were ASME Class 1 and ,

2 pipe weldments selected from the SI, MS, PZR, RHR, FW systems and steam generator ,

"C." Inspections were performed specifically to identify any cracks or linear indications, gouges, leakage, arc strikes with craters, or corrosion, which may infringe upon the i minimum pipe wall thickness and modifications to piping or components. Mirrors, flash l lights and weld gauges were uwd to aid in the inspection and evaluatio :

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Results: The welds examined were ground for inservice inspection prior to surface and volumetric examinations. The welding and overall workmanship inspected was acceptabl l The inspection reports of the licensee reflected the as-found conditions. No deviations were identifie l Liauid Penetrant Examination (PT) (57060)  ;

Fourteen (14) safety-related pipe weldments and adjacent base material (1/2 inch on either side of the weld) were examined using the visible dye, solvent removable method per NRC Procedure NDE-9, Rev.1, in conjunction with Site Procedure FNP-0-NDE-157.4, Rev. Included in this inspection were ASME Class 1 and 2 stainless and carbon steel pipe weldments selected from the SI, RHR, and FW system :

Results: The surface areas examined were adequately prepared for the examination. The ,

licensee recorded the same relevant indications noted by the NRC. No rejectable indications l were identified, no deviations note l

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t Ultrasonic Examination (UT) (57080)

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Twenty (20) safety-related pipe weldments were ultrasonically examined using NRC Procedure NDE-1, Rev.1, in conjunction with Site Procedure FNP-0-NDE-100.31, Rev. O, i i

and associated isometric drawings and ultrasonic inspection repons. Included in this examination were ASME Class 1 and 2 pipe weldments selected from the RH, SI, FW systems and steam generator "C." To obtain the greatest possible repeatability in performing the NRC independent evaluations, the examinations were performed utilizing ultrasonic units, transducers and cables that matched, as closely as possible, those used by the licensee. A distance amplitude correction curve was established utilizing Farley' calibration standard t Results: The ultrasonic examinations performed by the NRC closely matched within expected variations for this method, those of the licensees. However, for the ultrasonic ,

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examination on the feedwater pipe-to-nozzle weld, APR2-4350-22, the NRC found a t

questionable indication. A scanning level of 12 decibels (dB) over the reference level was used to obtain a 5% full screen height (FSH) inside diameter (ID) roll. At that scanning level, the indication had an amplitude of 40% FSH. Further investigation was required to ,

determine the nature of the indication. Different methods of crack detection and sizing,45*

shear wave, 60 shear wave,45* refracted longitudinal wave and a 30-70-70 sizing technique 4 (WSY-70 ), were employed to explore the NRC's finding. Due to the UT signal  ;

i characteristics, the NRC determined the indication to be a 2" long thermal fatigue crack with approximately a 0.3" throughwall dimensio The licensee was made aware of the indication by the NRC. The NRC performed a ,

verification examination with the licensee to assure the licensee understood the location and characteristics of the indication. The licensee performed several examinations to characterize >

the indication, including radiography, automated ultrasonic examination, P-Scan ultrasonic i examination and different angles of manual examination techniques. Based on the results of f

their examination and the EPRI Ray Trace plot of the ultrasonic beam, the licensee concluded the indication is root undercut as opposed to being a thermal fatigue crac i The feedwater nozzle examination is an augmented examination which is planned to be

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performed at Farley during every refueling outage. Subsequent to the inspection exit meeting, the licensee provided to the Region Il staff information on the operation of the auxiliary feedwater/feedwater system operation that shows the driving force for thermal fatigue crack growth to be relatively low due to the normally low auxiliary feedwater flow rate. Additionally, the fracture mechanics analysis provided at that time indicates that a '

crack of the size noted above, if it were present, would not propagate to unacceptable dimensions prior to the next refuel outage. The determination of whether the subject indication is a crack or a root undercut will be better understood at the next UT examination of the area. This is an unresolved item pending results of the UT examination of weld APR2-4350-22 at the next refuel outage (50-364/93-400-01). <

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Radiocrachic Examination (RT) (57090)

The NRC performed radiography, on FW weld APR2-4350-22, in an attempt to arbitrate the ultrasonic indication discussed above. The procedure and technique used to make the radiographs were in accordance with NRC Procedure NDE-10, Rev.1. The original construction radiographs were reviewed to compare the present condition to the preservice ;

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Results: The licensees radiography of the same area compared favorably to the NRC's i radiographs. The original radiographs were retrieved without difficulty and also compared favorably with the NRC's radiograph. The new radiographs neither confirmed nor refuted

the presence of a crac Erosion / Corrosion (E/C) (49001)

Concerns regarding erosion / corrosion (E/C) in balance of plant piping systems has ,

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heightened as a result of the December 9,1986 feedwater piping line rupture which occurred at Surry. This event was the subject of NRC Information Notice 86-106, issued December 16,1986, and its supplement issued on February 13, 198 The licensee's actions with regard to the detection of erosion / corrosion in plant components were reviewed with respect to NUREG-1344, " Erosion / Corrosion Induced Pipe Wall Thinning in U. S. Nuclear Power Plants," dated April 1989, Generic Letter 89-08 issued May 2,1989, and NUMARC Technical Subcommittee Working Group on Piping and Erosion / Corrosion Summary Report, dated June 11, 1987. Site Procedure FNP-2-M-063 provides the E/C program, plan and inspection summary. This is applicable to components in the feedwater, extraction steam, crossunder piping, heater and main steam reheater drain The basis for the Farley E/C program is the guidance of EPRI Report NP-3944, related '

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EPRI reports and industry input. Procedure FNP-2-ETP-4260 outlines the implementation of the E/C ultrasonic testing including establishing NDE reference points for E/C examinations :

and analysis of data. The areas reviewed by the NRC were system selection, component selection, ultrasonic inspection, and data evaluatio .

c The selection of systems to be analyzed and examined by the licensee were based on the systems susceptibility to E/C including flow velocity, steam quality, material of the ,

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components and system temperature. The components selected for independent ultrasonic inspection (UT) by the NRC include those with a high predicted erosion rate based on i

Keller's equation, similar components in sister trains, components in close proximity with those showing wear, industry experience, random sampling, components susceptible to ,

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cavitation and components which visually show outside erosio :

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Components selected for inspection by the licensee are ultrasonically inspected for thicknes ,

Other nondestructive methods including visual examination (VT) are utilized with UT or when UT is not applicable. A grid system in accordance with FNP2-ETP-4260 is used for the ultrasonic (UT) inspection. Each grid is scanned 100% with the lowest reading in each grid recorded. Five components were independently ultrasonically inspected by the NRC to determine the accuracy of the licensees' measurement The licensee's ultrasonic data is taken by a certified level II UT technician and the recorded data is evaluated by a certified level III UT examiner for accuracy and wall thickness. All data which is below the allowable wall thickness, as determined by the cognizant engineer, is evaluated and dispositioned prior to plant restart. The inspection information is the basis for the fm' al 90 day inspection report that includes a data base revision, repair / replacement recommendations and inspection recommendations for the next outag :

Results: Farley Nuclear Plant has an E/C program that includes application of NRC and EPRI guidance and industry experiences. The methods used by Farley personnel to select and l evaluate components are conclusive. The component inspections performed by the NRC closely matched those of the license .0 REVIEW OF SITE NDE PROCEDURES AND MANUALS (73052) )

Before a license's program of inspection is used, it must be approved by the Nuclear Regulatory Commission under the authority embodied in 10 CFR 50.55a (g) (4) (iv). The ,

required inspections are detailed in the American Society of Mechanical Engineers (ASME) i Boiler and Pressure Vessel Code,Section XI, for Inservice Inspection as embraced in 10 CFR 50.55a (b). For any inspection program, the code edition and addenda used is determined in accordance with the requirements of 10 CFR 50.55a (g).

Farley Nuclear Plant, Unit 2 submitted their second interval inservice,1st period, 2nd outage !

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inspection program to the NRC on July 12, 1993. The items and areas planned to be examined in this program, are in accordance with the Plant Technical Specifications Section 4.0.5. The 1983 Edition of Section XI of the (ASME) Boiler and Pressure Vessel Code up to and including Summer,1983 Addenda (83S83) is the applicable Code editio The following licensee procedures were reviewed for compliance to the applicable codes, standards and specification i i

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Procedure Title Number / Revision P_ ale Manual Ultrasonic Examination FNP-0-NDE-157.12 3/24/89 I of Welds Revision 0 l Manual Ultrasonic Examination FNP-0-NDE-12 8/3/93 l of Welds Revision 5 Manual Ultrasonic Examination FNP-0-NDE-100.34 9/24/93 of Welds in Vessels Revision 0 l 1 Manual Ultrasonic Examination FNP-0-NDE-100.31 9/24/93 j of Full Penetration Welds Revision 0 l

l Manual Ultrasonic Examination FNP-0-NDE-15 /7/90 l of Welds in Vessels Revision 3

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i Manual Ultrasonic Examination FNP-0-NDE-157.14 9/7/90 of Welds in Cast Stainless Revision 4 Steel Pipe f

Manual Ultrasonic Examination FNP-0-NDE-157.17 3/20/91 of Inner Radius Corners Revision 4 i

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Visual Examination VT-1 FNP-0-NDE-15 /7/90 Revision 2 i

Visual Examination VT-3 FNP-0-NDE-157.16 9/7/90 )

Revision 2 Liquid Penetrant Examination FNP-0-NDE-15 /7//90

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Revision 2 Magnetic Particle Examination FNP-0-NDE-157.11 10/18/90 Revision 3  ;

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Preservice and Inservice FNP-0-NDE-15 /28/90

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Examination Documentation

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Erosion / Corrosion Program FNP-2-M-063 5/4/93 Revision 3 Erosion /Corrosica Ultrasonic FNP-2-ETP-4260 10/3/93-

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Inspection Revision 1 i

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9 Results: Because the reports of NDE evaluations used two numbers for procedures (the vendors and the licensees), the tracking was confusing. Procedures were found to meet the intent of the referenced codes. No violations were identifie .0 MANAGEMENT MEETINGS Licensee management was informed of the scope and purpose of the inspection at the entrance meeting on October 4,1993. The fmdings of the inspection were discussed with the licensee representatives during the course of the inspection and presented to licensee management at tiie exit meeting October 14, 1993. The licensee did not indicate that proprietary information was involved within the scope of this inspection, nor did the licensee object to any of the findings of the inspection. The following individuals were contacted:

Southern Nuclear Comnany (Farley)

• B. Badham ISI Coordinator M. Mitchell H.P. Supervisor

• C. Nesbitt Operations Manager J. Odom Operation Unit Supervisor

• L. Stinson Operations M. Coleman Plant Modifications Manager

• R. Yance Manager Systems Performance

• S. Casey Suppon Supervisor S. Norman Mechanical Maintenance

• W. Bayne Quality Assurance Supervisor C. Buck Technical Manager T. Livingston Engineering Supervisor J. Hale Environmental Supervisor

• L. McClain NDE Level III

• R. Davis ITS T. Styers ITS U.S. Nuclear Regulatory Commission W. Kliensorge RII M. Scott RII, Farley Resident C. Miller Acting Deputy Director, Reactor Safety, RI

• Denote those attending entrance and exit meetin The inspectors also contacted other administrative and technical personnel during the inspectio . .

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l 10 COMPONENTS INSPECTED Following is a list of the components inspected by the NRC NDE Mobile 12boratory at the Farley Nuclear Plant:

NRC NDE MOBILE LABORATORY TABLE 1 WELD ID. N SYS NONDESTRUCTIVE TFSF Silt. #1 OR OR LIN CL RT UT Pr MT VT ACC REj ISO / DRAWING f

APR-2-4150-19 FW l X X X APR-2-4250-15 FW 1 X X X APR-2-4350-22 FW 1 X X X X APR-2-3300-5 SG 1 X X X APR-1-4111-1 Rl!R 1 X X X X APR-1-4111-2 RHR 1 X X X X APR-1-4111-3 RHR 1 X X X X i APR-1-4111-4 RHR 1 X X X X APR-1-4111-5 RHR 1 X X X X APR-1-4109-8 SIS 1 X X X X APR-1-4109-9 SIS 1 X X X X APR-1-4109-10 SIS 1 X X X .X APR-2-2210-1-12 FW 2 X X X APR-2-2212-1-51 MS 2 X X X ,

APR-1-4107-1 PZR 1 X X X X APR-1-4107-2 PZR 1 X X X X l

APR-1-4107-3 PZR 1 X X X X APR-1-4107-4 PZR 1 X X X X APR-1-4107-5 PZR 1 X X X X APR-1-4107-6 PZR 1 X X X X

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NRC INDEPENDENT MEASUREMENTS PROGRAM IIANGER/ SUPPORTS TABLE 2 IDENTIFICATION SYS CL ACC REJ COMMENTS APRI-4104-2SI-R134 SIS 1 X APR1-41M-2SI-R121 SIS 1 X APRI-4104-2SI-R120 SIS 1 X APR1-4104-2SI-R119 SIS 1 X APRI-4104-2SI-Ril8 SIS 1 X APRI-4104-2SI-Ril7 SIS I X APRI-41M-2SI-Ril6 SIS 1 X APRI-4104-2SI-Rt 15 SIS 1 X -

APRI-4101-2RH-R98 RHR 1 X APR1-4101-2RH-R99 RHR 1 X APRI-4101-2RH-R600 RHR 1 X