ML20153D652

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Jm Farley Unit 1 Interval 2,Period 1,Outage 1 Inservice Insp of Outlet Nozzle to Shell Weld 21:Recordable Indications Which Exceed Allowable Stds of ASME Section XI
ML20153D652
Person / Time
Site: Farley Southern Nuclear icon.png
Issue date: 05/31/1988
From: Adamonis D, Bond C, Kurek D
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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Shared Package
ML20153D649 List:
References
MT-SME-224, NUDOCS 8805090234
Download: ML20153D652 (47)
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{{#Wiki_filter:. MT-SME-224 EllCLOSURE 1 J. M. FARLEY UNIT 1 INTERVAL 2, PERIOD 1, OUTAGE 1 INSERVICE INSPECTION OF THE OUTLET N0ZZLE TO SHELL WELD #21: RECORDABLE INDICATIONS VMICH EXCEED THE ALLOWABLE STANDARDS OF ASME SECTION XI MAY 1988 D. KUREK W. H. BAMFORD R. D. RISHEL Verified by: C/ /M C. B. Bond Approved -=_1 D. C. Adamonis, Manager Materials Technology WESTINGHOUSE ELECTRIC CORPORATION P. O. BOX 2728 PITTSBURGH, PA 15230 m i. EBR5 X88M 888886 ~ o"

TABLE OF CONTENTS Section Page 1.0

SUMMARY

1-1 2.0 NONDESTRUCTIVE EXAMINATION RESULTS 2-1 2.1 Standard Inservice Inspection - 1988 Examinations 2-1 2.2 Comparison of Post Wechanized Inspection Results - 2-2 Preservice, 1984, and 1988 Examinations 2.3 Review of Fabrication Ultrasonic Test Reports 2-3 2.4 Review of Construction Radiographs 2-4 2.5 Supplemental Examinations with UDRPS 2-4 2.6 Indication Analysis 2-10 2.7 References 2-10 3.0 FRACTURE EVALUATION 3-1 3.1 Introduction 3-1 3.2 Fracture Properties 3-1 3.3 Irradiation Effects 3-1 3.4 Transients Evaluated 3-2 3.5 Treatment of Low Temperature Overpressurization Transients 3-2 3.6 Stress Corrosion Cracking Susceptibility 3-10 3.7 Results of Fracture Evaluation 3-11 3.8 References 3-12 'l0

O SECTION 1.0

SUMMARY

During the second interval, period 1 inservice examinations of the J. M. Farley Unit I reactor vessel outlet nozzle to shell welds performed in April 1988, two recordable indications were found to exceed the acceptance standards in table IWB-3512-1 of the ASME Code Section XI,1983 Edition with addenda up ) to and including the 1983 Summer Addenda. These indications were found using a zero degree, 2.25 MHz,1-1/2 inches diameter longitudinal wave transducer from the nozzle bore. A comprehensive evaluation of these indications was performed. This evalua-tion included a comparison of ultrasonic test results with past mechanized examinations and shop fabrication NDE examinations (manual ultrasonic examina-tions and radiography), and supplemental ultrasonic examinations using the Dynacon Ultrasonic Data Recording and Processing System (UDRPS). This evaluation concluded that these indications, identified as 3A and 22A, are embedded reflectors located at or near the weld / nozzle forging fusion line, that were deposited during the fabrication welding prt. cess. Their ultrasonic signatures using the UDRPS imaging system suggest rounded volumetric flaws such as slag or porosity. The 50% DAC sizing dimensions measured for these indications are a through-wall (2a) value of 1.32 inches and a length (t) value of 0.74 inches for indication 3A, and a through-wall (2a) value of 1.44 inches and a length (t) value of 0.64 inches for indication 22A. Using the rules of IWB-3600 and the guidelines of Appendix A, from the ASME Code Section XI,1983 Edition with addenda up to and including the 1983 Summer Addenda, these indications were subjected to a fracture analysis, and found to be acceptable. mwe.muae 13

SECTION 2.0 NONDESTRUCTIVE EXAMINATION RESULTS 2.1 Standard Inservice Inspection - 1988 Examinations During the April 1988 reactor vessel inservice inspection (interval 2, period s

1) two indications that exceeded the 50% DAC recording level and the acceptance standards of the ASME Code Section XI 1983 Edition up to and including the Summer 1983 Addenda were found.

These indications are located in the outlet nozzle-to-shell weld designated as weld #21. This weld is shown in the elevation and plan views of figures 2.1 and 2.2, respectively. . These indications were found using the Westinghouse Remote Inservice Inspection Tool with the standard 40-month array plate. The transducer that detected these indications was identified as TR6. It is configured to produce zero degree longitudinal waves from the nozzle bore. This particular transducer was a 2.25 MHz,1-1/2 inches diameter, pulse-echo unit. The transducer was calibrated in accordance with Section XI requirements using 3/8 inch diameter side-drilled holes at metal travel depths of 9-1/2 and 11-3/4 inches. A schematic of this arrangement is provided in figure 2.3. The two indications of interest have been identified as 3A and 22A. Indication 3A is located at approximately 124 degrees clockwise from the top center of the nozzle when viewed from the reactor vessel centerline. It achieved a maximum amplitude response of 100% DAC. Its measured through-wall (2a) dimension using 50% DAC sizing techniques is 1.32 inches. This dimension is parallel to the nozzle bore and therefore perpendicular to the inner diameter surface of the reactor vessel. Itsmeasuredlength(t) dimension using 50% DAC sizing techniques is 0.74 inch. This length dimension is parallel to the weld and therefore along a fixed radius around the nozzle bore. Table 2.1 includes the pertinent information concerning indication 3A. Figure 2.4 shows the location of indication 3A with respect to the nozzle azimuth as well as its location within the outlet nozzle-to-shell weld #21. It appears from this scaled figure that indication 3A lies along the nozzle forging fusion line of the weld and closer to the vessel outer diameter surface than to the vessel inner diameter surface. 2-1 l

w Indication 22A is located at approximately 100 degrees clockwise from the top center of the nozzle when viewed from the reactor vessel centerline. It had a maximum amplitude response of 100% DAC and has a measured through-wall (2a) i dimension'using 50% DAC sizing' techniques of 1.44 inches. This dimension is parallel to the nozzle bore and therefore perpendicular to the inner diameter j surface of the reactor vessel. Its measured length (t) using the ASNE Code specified 50% DAC sizing techniques is 0.64 inch. This length dimension is ] parallel to the weld and therefore along a fixed radius around the nozzle i bore. Table 2.1 includes 'this pertinent information. Figure 2.5 shows the location of indication 22A with respect to the nozzle azimuth as well as its location within the nozzle-to-shell weld #21. As with indication 3A, indication 22A appears to be located along the nozzle forging fusion line of the weld but closer to the center of the weld. Using the flaw indication evaluation rules of IWA-3000 and the acceptance standards for flaws indications of IWB-3000 (specifically table IWB-3512-1) it was determined that indication 3A has an actual a/t value of 7.3% compared to an allowable a/t of 6.5%, and indication 22A has an actual a/t value of 8.0% compared to an allowable a/t of 6.5%. Both indications exceed the allowables. i Table 2.2 provides the information used to calculate these values. 2.2 Comparison of Past Mechanized Inspection Results - Preservice.1984, and 1988 Examinations \\ In an effort to determine the history of these indications past mechanized inspection results related to the outlet nozzle-to-shell weld #21 were reviewed. These inspections included the preservice inspections performed in February 1977, the first interval inservice inspections in 1984, and the i second interval inservice inspections of 1988. l In the remote preservice examinations performed in February 1977 no recordable i indications were noted. These examinations included a O degree longitudinal I wave immersion bore examination using a 2.25 MHz, 1-1/2 inches diameter, pulse-echo transducer. The recording level, however, was 100% DAC, consistent with the applicable ASME Code Section XI of that time. = * * " " 2-2

The first interval inservice examinations performed in 1984 included the same zero degree longitudinal wave immersion bore examination with a similar transducer but with a recording level of 50% DAC. Four indications were found with this required scan as well as two others detected using an examination scheme using 16 degrees refracted longitudinal waves from the nozzle bore. Specifics on these indications are provided in table 2.3. In the second interval inservice exan,lnations performed in 1988 six indica-tions exceeded the 50% DAC recording levels and required evaluation. These examinations duplicated the first interval examinations in terms of technique (0 degree longitudinal wave, 2.25 MHz,1-1/2 inches diameter, pulse-echo transducer) and recording level (50% DAC). Three of these recorded indica-tions match the first interval results allowing tolerances for locations betweeninspectionsandamplitudediscrepanciesof+/-3dB(table 2.3). The other three of the 1984 indications were found but were not recorded because their maximum amplitude did not exceed 50% DAC. Three new indications were found. These included two with the 20 degree longitudinal wave examination from the nozzle bore and one with the O degree longitudinal wave examination from the nozzle bore. The two new 20 degree longitudinal wave indications can be explained by the change in the examination angle from 16 degrees to 20 degrees. These two indications, however, were low amplitude and were determined to be acceptable per IWB-3512-1. The new 0 degree longitudinal wave indication could be the result of a slight change in ultrasonic parameters with respect to this particular reflector. Table 2.3 shows the specifics of these indications with respect to the other previous examinations. 2.3 Review of Fabrication Ultrasonic Test Reports In 1973 the J. M. Farley Unit I reactor vessel was examined by manual ultra-sonic examination techniques for acceptance to ASME Code Section III paragraph N-625, 1968 Edition including addenda up to and including Summer 1970. Reports from the outlet nozzle to shell weld #21 (CEf1-897E) were reviewed in an effort to establish whether or not a correlation exists with the 1988 remote inservice ultrasonic data. These examinations consisted of 0 degree me.omone " 2-3

longitudinal wave, 45 degree shear wave and 60 degree shear wave manual contact techniques with a 100% DAC recording level. These examinations were calibrated on calibration blocks different than those used for the inservice inspections. No recordable indications were noted. 2.4 Review of Construction Radiographs Construction radiographs of the J. M. Farley Unit 1 reactor vessel outlet nozzle to upper shell weld #21 (CEf1-897E) were reviewed to establish whether or not a correlation exists with the 1988 ultrasonic examination data. The radiographic technique for the nozzle to shell weld #21 specified the use . of Kodak AA film in a double loaded cassette located on the inner diameter with a high energy X-ray source located on the outside diameter. The incident X-ray beam was angled from the normal so as to include as much of the nozzle to shell weld volume as practical. No relevant flaw-type images were noticeable on the construction radiographs of the outlet nozzle to shell weld #21. 2.5 Supplemental Examinations with UDRPS In order to obtain better information regarding the nature and size of the ultrasonic indication (3A and 22A) it was decided to utilize the Dynacon UltrasonicDataRecordingandProcessingSystem(UDRPS)withtheconventional inservice inspection transducers. The UDRPS system is a known automated data recording and processing system which has the capability of recording, storing, processing, and imaging ultrasonic test data. It allows for more extensive recording of data, better visualization of examinstion data through the use of color-coded images, more flexible manipulation of data, more consistent examination quality, and archival retrieval of past examinations for comparison purposes. 2-4

y s The best use of the UDRPS data is the ability to observe secondary re@ases and their relationship to the primary signals from the indications. This aids in the characterization of the reflectors as well as potentially providing more accurate sizing information. Characterization is defined as "the determination of whether a valid indica-tion originates from a volumetric or planar type defect". Generally, the use of supplemental straight beam and angle beam techniques provide for the veri-fication of a volumetric type flaw, i.e. slag, porosity, since a relatively strong reflection should occur from both. Planar flaws should reflect little or no energy to a straight beam transducer. Another supplemental characterization technique is based on satellite pulse observation technique (SPOT) principles [ reference 2-1). SPOT relies on the observation of a doublet signal emanating from a volumetric defect. This doublet consists of a strong specularly reflected signal, followed by a weak, synchronous satellite pulse response. This satellite pulse is created by a portion of the sound beam propagating around the circumference of a rounded type of reflector and being reradiated back to the receiver transducer. Synchronous means that when the specularly reflected signal peaks the associated satellite pulse signal should also peak with the satellite pulse lagging in arrival time. Therefore these two peaks should occur in the same A-scan. On a system such as UDRPS two relatively close parallel images one behind the other would be indicative of synchronous signals and therefore a rounded volumetric type of defect. For planar flaws, SPOT also relies on the observation of a doublet signal but these signals are asynchronous in nature, in this case the satellite responses are created by a portion of the cound beam being reradiated from a planar flaw extremity back to the receiver transducer. Since the extremities of planar flaws are separated in position the peaks of each extremity would not occur in the same A-scan. On a system such as UDRPS two parallel images but shifted in position would be indicative of asynchronous signals and therefore a planar type of defect. ano.** " 2-5

I UDRPS was also used to determine the reflector sizes using amplitude drop sizing methods. The UDRPS system, however, has the same fallacies as conven-j i tional ultrasonic examination techniques using amplitude drop sizing methods. For small flaws it will still provide estimated sizes more commensurate with the beam size of the transducer rather than the size of the flaw, assuming as in most cases that the beam size is greater than the size of the flaw (references 2-1 thru 2-5). With this in mind, for the indications in the outlet nozzle-to-sbell weld #21, i a sizing methodology known as - 6 dB amplitude drop or half maximum technique was applied. This methodology was applied because overall (on a defect matrix consisting of volumetric and planar type flaws) it has been shown to provide I the more accurate results when compared to other amplitude-based techniques i such as 50% DAC, 20% DAC, and 20% DAC with beam spread correction (reference l 2-6). I Indications 3A and 22A found during the conventional inservice inspection were re-examined using UDRPS. They were found using the Westinghouse 40-month array plate. The transducer that detected these indications is a 2.25 MHz, i 1-1/2 inches diameter, zero degree longitudinal wave unit identified as TR6. j l This same transducer was used in the UDRPS scans. For the UDRPS examinations l three scans were performed on each indication. Each of the scans was I conducted in a raster fashion using the Y-axis of the mechanized scanner (toward and away from the reactor vessel centerline) as the scanning axis and using the B-axis of the mechanized scanner (around the nozzle bore) as the f indexing axis. Each index was 0.5 degrees around the nozzle or a 0.22 inch 1 increment at the location of the indications. The first UDRPS scan was [ performed at the sensitivity required to achieve a maximum amplitude of 80% i full screen height from the indication of interest. The remaining two scans were performed at sensitivities 10 dB and 16 dB above the first scan level, i f These higher sensitivity scans were performed to try to distinguish secondary l responses which could determine the nature of the indications. The lower sensitivity scan was performed to enable a - 6 dB drop sizing methodology to be applied. L l Y l t j i

=== uu ie 2-6 b i l I

F The results of these examinations can best be observed in figures 2.6 through 2.11. A brief explanation of each of these figures is provided below: Figure 2.6: Indication 3A-Cross-sectionalView(TransducerTR6-Low { i Sensitivity Scan), UDRPS Data e This image is of the scan line that showed the maximum response from indication 3A (124.1 degrees nozzle azimuth). As a means of under-standing the geometry of the scan, estimated positions of the shell, weld l and nozzle have been shown. The scans were established such that a I response from the nozzle bore outer diameter could be observed in order { l to determine the relative position of the indication. As can be seen j indication 3A appears to be at or near the weld / nozzle forging fusion line and clearly embedded within the weld. Using - 6 dB drop sizing, i.e. points where the amplitude drops to half the maximum value, the 2a l dimension of this indication is determined to be 1.2 inches. This dimension is shown on the figure. The minimum distance from the half i maximum extremity point of this indication to the intersection of the weld taper and the nozzle bore outer diameter is approximately 4.5 inches. l I j Figure 2.7 : Indication 3A - Linear Extent (Transducer TR6 - Low SensitivityScan),UDRPSData i l The series of images shown in this figure displays the linear extent of l indication 3A without any amplitude drop type sizing. Application of - 6 l l dB drop sizing results in a linear extent ranging from 123.6 to 126.1 l l degrees or 6 increments. Each increment is 0.5 degrees therefore the f l indication extents over 3 degrees. Since the weld fusion line is 51 f inches diameter from the nozzle centerline, this establishes a conver-l l sion factor of 0.445 inch / degree. Therefore indication 3A by - 6 dB drop sizing measures 1.34 inches in length. t i k 1 I 4 ) I -4 2-7 i i i t

a. l l i figure 2.8 : Indication 3A - Secondary Images (Transducer TR6 - High Sensitivity Scan +10 dB) UDRPS Data This image along the scan line at 124.1 degrees shows a saturated response from indication 3A as well as a weak synchronous trailing secondary response approximately 3 microseconds behind the primary response from the indication. Both these responses are identified on the figure. Synchronous, trailing secondary responses are an indicator of rounded volumetric r6fle'etors such as slag or porosity. Figure 2.9 : Indication 22A - Cross-sectional View (Transducer TR6 - Low Sensitivity Scan), UDRPS Data This image is of the scan line which showed the maximum 2a dimension for indication 22A at half maximum (- 6 dB) amplitude points (100.0 degrees nozzleazimuth). Estimated positions of the shell, weld and nozzle are shown for clarity. Indication 22A is actually two reflectors approxima-tely 0.6 inch apart (peak to peak) located at or near the weld / nozzle forging fusion line and embedded within the weld. Using - 6 dB drop sizing and ASME Code 1983 Edition with addenda up to and including the 1983 Summer Addenda Section XI proximity rules the 2a dimension for indication 22A is determined to be 1.58 inches. This dimension is shown on the figure. The minimum distance from the half maximur. extremity point of this indication to the intersection of the weld tsper and the nozzle bore outer aiameter is approximately 5.5 inches. Figure 2.10: Indication 22A - Linear Extent (Transducer TR6 - Low Sensitivity Scan), UDRPS Data The series of images shown on this figure display the linear extent of indication 22A without any amplitude drop type sizing. The presence of other various reflectors with lower amplitudes is clearly evident. Using - 6 dB drop sizing indication 22A can be seen ranging from 102.0 to 102.3 degrees, from 99.0 to 101.0 degrees and at 97.5 degrees. Using ASME Code am." " 2-8

proximity rules indication 22A extends from 97.5 to 102.5 degrees or 11 increments. Each increment is 0.5 degrees therefore the indication extents approximately 5.5 degrees. This converts to a length of 2.45 inches (.445 inch /degreeattheapproximatelocationofindication). Figure 2.11: Indication 22A - Secondary Images (Transducer TR6 - High Sensitivity Scan +10 dB) UDRPS Data This image shows a saturated response from indication 22A as well as a weak synchronous trailing secondary response approximately 3 to 3.5 microseconds behind the primary indication 22A response. Both responses are identified on the figure. Synchronous trailing secondary responses are indicative of rounded volumetric type flaws such as slag or porosity. Using UDRPS and - 6 dB amplitude drop sizing with the ASME Code 1983 Edition with addenda up to and including the 1983 Summer Addenda Section XI rules for flaw evaluation, indication 3A is a subsurface flaw indication having a 2a dimension of 1.20 inches and a length of 1.34 inches, and indica-tion 22A is a combination of at least two subsurface flaw indications having a combined 2a dimension of 1.58 inches and a combined length of 2.45 inches. Both 4 indications have evidence of synchronous trailing secondary responses which are indicative of rounded volumetric type reflectors such as porosity and slag. Using the flaw indication evaluation rules of IWA-3000, the acceptance standards for flaw indications of IWB-3000 (specifically table IWB-3512-1), and the dimensions of the indications from the supplemental examinations with UDRPS (-6 dB drop sizing), it is determined that indication 3A has an actual a/t value of 6.7% compared to an allowable a/t of 5.8%, and indication 22A has l an actual a/t value of 8.8% compared to an allowable a/t of 4.3%. Both indications exceed the allowables of table IWB-3512-1. Table 2.4 provides the information used to calculate these values.

== - 2-9

o 2.6 Indication Analysis From this comprehensive evaluation effort which entailed a review of past NDE results, the results of the conventional 1988 inservice inspections, and the supplemental examinations using UDRPS it appears that indications 3A and 22A are rounded volumetric type reflectors that are clearly subsurface in nature. They appear to be located at or near the weld / nozzle forging interface (figures 2.4 and 2.5) In terms of their through-wall and length dimensions two different measurements have been taken, one using 50% DAC sizing methods and taken manually, and the other using - 6 dB drop sizing methods and taken with UDRPS. These combined dimensions are given in table 2.5. While amplitude-drop sizing methodologies are not considered that accurate the - 6 dB drop methodology has been shown to provide the more accurate results when compared to 50% DAC, 20% DAC, and 20% DAC with beam spread correction sizing methods [ reference 2-1). 2.7 References 2-1 Gruber, G. J., G. J. Hendrix, and W. R. Schick. "Characterization of Flaws in Piping Welds Using Satellite Pulses", Materials Evaluation, Volume 42, April 1984, pp.426-432. 2-2 Cook, R. V., P. J. Latimer, and R. W. McClung. Flaw Measurement Using Ultrasonics in Thick Pressure Vessel Steel, Final Report on Contract No. W-7405-eng-26, prepared by Oak Ridge National Laboratory for the U. S. Nuclear Regulatory Commission, August 1982, Oak Ridge, TN. 2-3 Doctor, S. R., et al. "Effectiveness of U.S. Inservice Inspection Techniques - A Round Robin Test", Proceedings of Specialist Meeting on Defect Detection and Sizing, ispra, Italy, May 3-6, 1983. 2-4.lessop, T. J., P. J. Mudge, and J. D. Harrison. Ultrasonic Measurement of Weld Flaw Size, National Cooperative Highway Research Program Report 242, prepared for the Transportation Research Board by The Welding Institute, December 1981. an m a io 2-10

2-5 Mudge, P. J., and T. J. Jessop. "Size Measurement and Characterization of Wald Defects by Ultrasonic Testing : Findings of a Collaborative Prograrwe", Proceedings of NDE in Relation to Structural Integrity, Paris, France, August 24-25, 1981. 2-6 Willetts, A. J., F. V. Ammirato, and E. K. Kietzman, J. A. Jones Applied Research Company / EPRI NDE Center. Accuracy of Ultrasonic Flaw Sizing Techniques for Reactor Pressure Vessels, Draft Interim Report, EPRI RP 1570-2, March 1988. amaesum ie 2-11

a. TABLE 2.1

SUMMARY

OF INDICATIONS 3A AND 22A 0F THE DUTLET N0ZZLE TO SHELL WELD #21 IND. DETECTION DETECTION CALCULATED CALCULATED MEASURED MEASURED NO. TRANSDUCER ANGLE DEPTH FROM METAL PATH THROUGH-LENGTH EXAMINATION WALL SURFACE DIMENSION (2a) (1) 3A TR6 0 deg. L 10.88" 10.88" 1.32" 0.74" 22A TR6 0 deg. L 10.15" 10.15" 1.44" 0.64" 1 TABLE 2.2 EVALUATION OF INDICATIONS 3A AND 22A 0F THE OUTLET N0ZZLE TO SHELL WELD #21 IND. MEASURED MEASURED "S" TYPE OF ASPECT a/t ALLOWABLE NO. THROUGH-LENGTH VALUE* FLAW RATIO ACTUAL a/t WALL (2a) (1) 3A 1.32" 0.74" 1.4" subsurf. 0.5 7.3% 6.5% i 22A 1.44" 0.64" 3.5" subsurf. 0.5 8.0% 6.5% l i i "S" is measured using scaled plots in figure 2.4 and figure 2.5 and is equal to the minimum measured dimension from the 50% DAC extremity point to the 0.D. weld taper. The weld taper dimension is taken from f the design drawings. "t" is taken as 9.0 in. based on an average UT thickness of the nozzle shell minus a cladding thickness of approximately.25 in. From ASME Code 1983 Edition with addenda up to and including the 1983 Summer Addenda Section XI Table IWB-3512-1.

== is 2-12

p. TABLE 2.3 COMPARISON OF PAST MECHANIZED INSPECTIONS OF OUTLET N0ZZLE TO SHELL WELD #21 FIRST INTERVAL SECOND INTERVAL e' 1984 EXAMINATION RESULTS 1988 EXAMINATION RESULTS IND. ANG. R* B** AMPLITUDE IND. ANG. R* B** AMPLITUDE NO. (deg) (in)(counts) (%DAC) NO. (deg) (in)(counts) (% DAC) 1 16 82.5 4303 63 3 20 82.5 4263 56 2 16 82 30455 65 20 82 30420 20-50 max 1 0 84.25 39602 65 0 84 39603 50 max 2 0 83 34575 69 3A 0 83 34466 100 3 0 82.75 9709 100 4A 0 82.75 9723 65 4 0 82 32584 50 0 82 32351 50 max 1 20 83.5 34496 60 i 2 20 79 21938 68 22A 0 79 32040 100 "R" is the location of the indication with respect to the reactor vessel i centerline. It is defined as the radius and is estimated based on mechanized tool counts, t "B" is measured in counts. It is the remote inservice inspection tool's axis of motion around the nozzle bore. The conversion factor is 100 counts per degree.

      • Indication not recorded due to maximum amplitude not exceeding 50% DAC i

therefore no indication number given. j f t s i l answeseees to '3 t o1

l TABLE 2.4 i EVALUATION OF INDICATIONS 3A AND 22A 0F THE l' DUTLET N0ZZLE TO SHELL WELD #21 USING SUPPLEMENTAL ULTRASONIC EXAMINATION DATA FRON UDRPS IND. NEASURED MEASURED "S" TYPE OF ASPECT a/t ALLOWABLE NO. THROUGH-LENGTH VALUE* FLAW RATIO ACTUAL a/t WALL (2a) (1), 3A 1.20" 1.34" 1.4" subsurf. 0.44 6.7% 5.8% 22A 1.58" 2.45" 3.5" subsurf. 0.32 8.8% 4.3% "S" is measured using scaled plots in figure 2.4 and figure 2.5 and is equal to the minimum measured dimension from the 50% DAC extremity point to the 0.D. weld taper. The weld taper dimension is taken from the design drawings. "t" is taken as 9.0 in, based on an average UT thickness of the nozzle shell minus a cladding thickness of approximately.25 in.. From ASME Code 1983 Edition with addenda up to and including the 1983 Summer Addenda Section XI Table IWB-3512-1. 2-14

l TABLE 2.5 INDICATION ANALYSIS OF INDICATIONS 3A AND 22A l IN THE OUTLET N0ZZLE TO SHELL WELD #21 e ? IND. NEASURED NEASURED "S" TYPE OF r NO. THROUGH-LENGTH VALUE* FLAW CONNENTS i i WALL 4 (2a) (t) { i 1 l 3A 1.32" 0.74" 1.4" subsurf. manual 50% DAC sizing with j mechanized tool 22A 1.44" -0.64" 3.5" subsurf. manual 50% DAC sizing with mechanized tool 3A 1.20" 1.34" 1.4" subsurf. - 6 dB drop sizing with UDRPS I f 22A 1.58" 2.45" 3.5" subsurf. - 6 dB drop sizing with UDRPS "S" is measured using scaled plots in figure 2.4 and figure 2.5 and i is equal to the minimum measured dimension from the 50% DAC extremity i point to the 0.D. weld taper. The weld taper dimension is taken from i the design drawings. { i i l i k [ l i 4 i 1 i I 8"a **" " 2-15 i r

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SECTION 3.0 FRACTURE EVALUATION 3.1 Introduction Tho fracture evaluation is described in detail in reference [3-1), the Background and Technical Basis for the Handbook in Flaw Evaluation for the Joseph Farley Nuclear Plant Units 1 and 2 Reactor Vessel Beltline and Nozzle to Shell Welds. This section will highlight the key elements of the fracture anelysis, and the results will be presented. 3.2 Fracture Properties The fracture properties of the outlet nozzle to vessel weld region were taken from the ASMI Code Section XI Appendix A. The reference toughness curves are adjusted for the value of RTNDT of the material. Table 2-2 of reference (3-1] contains the values for the nozzle forgings and upper shell material. Specific test results are not available for the nozzle forgings so the RTNDT was estimated as 60'F, using the guideline of NUREG 0800 [3-2). Test results were likewise unavailable for the nozzle to shell weld material, but it was required to meet a minimum Charp energy of 30 ft lb at 10'F, so RTNDT = 10*F per the estimation procedure of NUREG-0800 (3-2). 3.3 Irradiation Effects The level of irradiation damage at the outlet nozzle to shell weld of the Joseph Farley Unit i reactor vessel is three orders of magnitude lower than at the core midplane. The end of life fluence for the vessel inner surface was calculated based on operation at 2652 Mwt for 32 EFPY and assuming that exposure for all cycles after 7 is at the same rate as that calculated for cycle 7 [3-3). The end of life fast neutron fluence at the vessel inner surface is 3.88 E19 N/SQCM based on the measurements of the most recent surveillance capsule. Reducing this fluence appropriately for the outlet nozzle to vessel weld, the applicable fluence is 3.88 E16 N/SOCM at the lowest point of the weld. The indications closest to this location was designated me.o.mowo 3-1

i 3A, at 120 degrees from the top of the nozzle, and 4A at 238 degrees from the nozzle top. These locations are approximately 10 inches above the lowermost point of the nozzle to shell weld, and this additional height results in a further reduction in fluence by a factor of 10. Therefore, the fluence applicable to the indications of interest here is 3.88 E15 at the vessel inner surface. This value would be further reduced by the fact that the indications are embedded. Clearly at these fluences irradiation damage is negligible. The fluence values used herein were taken from the most recent surveillance capsule test report, WCAP-11563, Sept.1987. [3-3). The fluences quoted above have been taken directly from figure 6-4 of reference (3-3), which is reproduced here as figure 3-1. The actual distance from the core mid plane to the lowest point of the nozzle-to shell weld is 102.93 inches, as shown in figure 3-2, and the indication at the lowest elevation is an additicnal 10.8 inches above this, for a total distance of 113.73 inches. 3.4 Transients Evaluated All applicable transients have been evaluated in developing the flaw evaluation charts, including pressurized thermal shock transients and all design transients as discussed in reference [3-1), section 2.1. In addition to these transients, low temperature overpressure transients w&re considered, as will be discussed below in section 3.5. 3.5 Treatment of Low Temperature Overpressurization Transients (LTOP) In this section, the frequency of occurrence of a significant low temperature overpressurization (LTOP) challenge to the Joseph M. Farley Plants (FP) is calculated. For this calculation, a probabilistic risk assessment (PRA) model was used. The model was based on event tree analysis, which defines the l possible event scenarios that may lead to a significant LTOP challenge for the nuclear plant. This work demonstrates that an LTOP event is of such a low probability that it may be classified as a faulted condition. i f 32 1 L-

F. 3.5.1 Model Assumptions The PRA model is constructed on the basis of the following assumptions: 1. The LTOP event is of concern if the primary coolant temperature is less than 350 degrees F. 2. There are written operational procedures for overpressurization mitigation: The operators are trained in mitigating the LTOP event. 3. During plant cooldown prior to reducing reactor coolant system temperature below the minimum temperature allowable for the inservice pressure test, valves 8702A/B and 8701A/B are opened to ellow any excess pressure to be vented through the Residual Heat Removal System Relief Valves (RHRSRV) 4. During cooldown, ths RHRSRV are placed in service between 350 and 310 degrees F. while a steam bubble remains in the pressurizer. After RHR is placed inservice at less than 350 degrees F., the reactor coolant system is cooled and taken solid after reaching 160 degrees F. 5. Normal operating procedures maximize the use of a pressurizer cushion (steam bubble) during periods of low temperature operation. A steam bubble is formed in the pressurizer at a cold leg temperature of approximate 160'F when the plant is being started up. It is collapsed at a cold leg temperature of approximately 160'F when the plant is being cooled down. This cushion substantially reduces the severity of some potential transients such as RCP-induced heat input and slows the rate of pressure rise for others. This provides reasonable assurance that most potential transients can be terminated by operator action before an overpressure condition exists. amo.co nio 3-3

6. Additional limitations placed upon plant operations include: a) When the reactor coolant system is not open to the atmosphere and the temperature of one or both reactor coolant system celd legs is less than or equal to 180 degrees F., no more than one high pressure safety injection (SI) pump shall be operabla. The second and third SI pump breakers shall be racked out. b) A reactor coolant pump (RCP) shall not be started when the reactor coolant system temperature is less than the minimum temperature for the inservice pressure test unless: l 1. There is a pressure absorbing volume in the pressurizer. 2. The secondary water temperature of each steam generator is less than 50 degrees F above the temperature of the l reactor coolant system. j 7. Several control room alarms have been provided. A seismic category 1 alare designed to the requirements of IEEE-279-1971 is in service i and alerts the operator if the RtiR isolation valves are not fully open when the RCS temperature is less than or equal to 350*F. Another alarm provides indication to the operator o' any overpressure transient occurring when the RCS pressure exceeds 425 j psig. l l 8. Above 180'F, the motor operated valves (MOVs) upstream of the safety l valves can close from either a spurious closure, or a falso high pressure input signal which can close both MOV trains. for a general surveillance time of these valves, a conservative period of 10 hours will be assumed. 9. Although the PORVs are available to the operator to mitigate an overpressure event, no credit will be taken for the PORY: and corresponding operator action (s) to reduce an LTOP event. acun ie 34

10. The failure rate for a sensor (2.8E-06/hr) includes all failure modes (i.e., incorrect /no signal and spurious signal). Although a spurious signal may make up a very small percent of this number, the conservative number of 2.8E-06/hr will be used for a spurious signal.

3.5.2 Event Tree Analysis Event tree analysis is use,d to model and quantify the progression and frequency of significant LTOP' challenges. The event tree considers the following six items in the progression of a significant LTOP challenge: 1. An LTOP precursor challenges the plant safety systems as an initiating event. This challenge may be due to inadvertent addition of heat or mass into the primary system, e.g. in the form of startup of a reactor coolant pump or a SI pump, or due to blockage of letdown lines. 2. Primary coolant temperature during the challenge is below 350 degrees F. If not, the LTOP event is not a concern. 3. The RHRSRV system is available. This system is available if one or two of the trains are available. 4. The pressurizer has bubble formation. If the bubble exists, operator action is possible even if the RHRSRV system fails. l 5, High pressure alarms and valve closure work. If they do, the operator can take manual action. Operator action is only credible if a bubble is present. Otherwise, the event progresses too fast for operator response. 6. Operator mitigates the event. This is credible if the bubble is present and the alarms work. l l l 35 l

3.5.3 Success Criteria The success criteria to avoid a significant LTOP event is: a) At least one mitigating system train is available. b) If both trains of the mitigation system fail, then there must be bubble in the primary and the alarm must work and the operator action must be successful. 3.5.4 Data Used Two sets of data were used for the calculation of the frequency of a significant LTOP event. The first set was a best estimate calculation; the second one was a conservative estimate calculation. 1. Initiating event frequency. (Event tree node OPC, in table 3-1) Since 1983, there have been 3 challenges for 2 plants. Thus the best estimate challenge frequency is calculated as f = 3 / (2

  • 5 years) = 0.3 per calendar year; per plant.

As a conservative approximately twice the above value will be used: 0.6/ calendar year. 2. Primary coolant temperature during the challenge is less than 350 degrees F. (Event tree node TMP) The probability of the temperature being in the range of concern is unknown. It will be taken as 0.9, assuming that challenge occurs 90 percent of the time in the temperature range of concern. The same value will be used for the conservative estimate. m e. c o.= to 3-6

3. Overpressure protection is available.(Event tree node OPS) The unavailability of each train upon demand will be calculated as follows: RHRSRV fails to open on demand 3.0E-04/ demand. RHRSRV is out-of service (1 day per year) 2.7E-03 1 of 2 motor operated valves spuriously close 2 (1.0E-07)(10) = 2.0E-06 Total per train: 3.0E-03/ train Probability of any one of the two trains being unavailable is 2

  • 3.0E-03

= 6E-03. Probability of both trains being unavailable is calculated as follows: q = random failures + 2

  • one train in maintenance and the second fails by random causes + failure of pressure sensors + both trains fail by common cause.

Assuming a beta factor of 0.10 for common cause failure, q = (3.0E-04)(3.0E-04) + 2 * (2.7E-03)(3.0E-04) + 2(2.8E-06)(10) 0.10*(3.10E-04) q = 8.7E-05. 4. The pressurizer has a bubble formation.(Event tree node WS0) It will be assumed that there is no bubble formed in 80 percent of the challenges. This is due to operational practices in which the bubble is formed after heat-up, and primary water being taken solid during cooldown. The same value will also be used for conservative estimate. -u.io 3-7

~ 5. High pressure and valve closure alarms work.(Event tree node ALR) It will be assumed that the failure of these a 4tms occurs once in hundred challenges. This is considered to be conservative. 6. Operator mitigates the event.(Event tree node OPE) The operator failure probability to respond to the alarm within a five minute period may be taken as 0.01, assuming that there are at least two operators in the control room and the alarm works; and there is a bubble formation in the primary, giving the operators a three to ten minute response time. However, as a conservative estimate and no credit is given for the PORVs, operator failure probability of 1.0 will be used. 3.5.5 Data Summary The following data were used to quantify the event tree for best estimate and conservative estimate scenarios for the frequency of a significant LTOP challenge per calendar year, per plant: l l BEST ESTIMATE CONSERVATIVE l Initiating event frequency 0.3 0.6 l l Primary temp. is less than 350 F. 0.90 0.90 Overpressure protection availability: Both trains are available 0.9939 0.9939 Only one train is available 6.0E-03 6.0E-03 l Both trains are unavailable 8.7E-05 8.7E-05 f 1 3-8 l l

BEST ESTIMATE CONSERVATIVE No bubble in the pressurizer 0.80 0.80 Alarm fails 0.01 0.01 Operator fails to mitigate Alarm works / bubble available 1.0 1.0 Alarm fails 1.0 1.0 Bubble not available 1.0 1.0 3.5.6 Calculations and Conclusions The event trees in tables 1 and 2 cre used to calculate the frequency of a significant LTOP event per calendar year per plant on the Farley site as follows: The best estimate freauency is 2.3E-05 per calendar year. The conservative estimate frequency is 4.7E-05 per calendar year. Based on the above results, the frequency of a significant LTOP event is very low for the nuclear units on the Farley Site. According to the event cate-gorization rules of ANS and the NRC Regulatory Guide 1.48, figure 3-3, this LTOP event is clearly classified as a faulted condition. This event was not a governing transient for the flaw evaluation charts because it is much less severe than the other favited conditions. It is of interest to compare the results of the work presented here with a study presently in progress, "Residual Heat Removal System Autoclosure l Interlock Removal Report for the Joseph M. Farley Nuclear Plant Units 1 and 2," WCAP-11746 (3-4). This report, which is yet to be published, is more 3-9 l l

r l detailed in showing the various plant states resulting from an LTOP initiating event, and indicates a frequency of high/significant overpressure (similar to the Turkey Point incident) of 3.0E-07/ shutdown year or 1.0E-07/ plant year. Thus, the results of this analysis (2.3 and 4.7E-05/ plant year) are very conservative relative to the more detailed assessment of reference (3-4). Both analyses indicate the LTOP event is not a governing transient for the flaw evaluation charts because it is much less severe than the other faulted conditions. 3.6 Stress Corrosion Cracking Susceptibility In evaluating flaws, all mechanisms of suberitical crack growth must be evaluated to ensure that proper safety margins are maintained during service. Stress corrosion cracking has been observed to occur in stainless steel in operating BWR piping systems; the discussion presented here is the technical basis for not considering this mechanism in the present analysis. For all Westinghouse plants, there is no history of stress corrosion cracking failure in the reactor coolant system loop. For stress corrosion cracking (SCC) to occur in piping, the following three conditions must exist simultane-ously: high te<sile stresses, a susceptible material, and a corrosive environment. Since some residual stresses and some degree of material suscep-tibility exist in any stainless steel material, the potential for stress corrosion is minimized by proper selection of materials immune to SCC as well as preventing the occurrence of a corrosive environment. The material specifica-tions consider compatibility with the system's operating environment (both internal and external) as well as other materials in the system, applicable ASME Code rules, fracture toughness, welding, fabrication, and processing. The environments known to increase the susceptibility of austenitic stainless steel to stress corrosion are oxygen, fluorides, chlorides, hydroxides, hydrogen peroxide, and reduced forms of sulfur (e.g., sulfides, sulfites, and thionates). Strict cleaning standards prior to operation and careful control of water chemistry during plant operation are used to prevent the occurrence neo.memae 3-10 [

of a corrosive environment. Prior to being put into service, the piping is cleaned internally and externally. During flushes and preoperational testing, water chemistry is controlled in accordance with written specifications. External cleaning for Class 1 stainless steel piping includes patch tests to monitor and control chloride and fluoride levels. For preoperational flushes, influent water chemistry is controlled. Requirements on chlorides, fluorides, conductivity, and pH are included in the acceptance criteria for the piping. During plant operation, the r'eactor coolant system (RCS) water chemistry is monitored and maintained within very specific limits. Contaminant concentra-tions are kept below the thresholds known to be conducive to stress corrosion cracking with the major water chemistry control standards being included in the plant operating procedures as a condition for plant operation. For example, during normal power operation, oxygen concentration in the RCS is expected to be less than 0.005 ppm by controlling charging flow chemistry and maintaining hydrogen in the reactor coolant at specified concentrations. Halogen concentrations are also stringently controlled by maintaining concen-trations of chlorides and fluorides within the specified limits. This is assured by controiling charging flow chemistry and specifying proper wetted surface materials. 3.7 Results of Fractu_re Evaluation The results of the fracture meenanics analysis performed for the outlet nozzle region of the Farley Unit i reactor vessel are shown graphically in figure 3-4. This figure is a flaw evaluation chart, and was taken directly from figure A-4.5 of reference 3-1. The indication characterization sizes from section 2 of the report are repeated here for convenience: INDICATION NO. 2a a/t i S 6/t 3A 1.32 0.086 0.74 1.4 0.242 3A(UDRPS) 1.20 0.067 1.34 1.4 0.223 22A 1.44 0.080 0.64 3.5 0.468 22AUDRPS) 1.58 0.088 2.45 3.5 0.476 m e. m s u to 3 11

These indications are plotted in figure 3-4, where it is seen that the indications are acceptable by a wide margin. 3.8 References

-1 Bamford, W. H., Balkey, K. R., and Lee, Y. S., "Background and Technical Basis for the Handbook on Flaw Evaluation for the Joseph M. Farley Nuclear Plant Units 1 and 2 Reactor Vessel Beltline and Nozzle-to-Shell Welds" Westinghouse Report WCAP-11763, April 1988.

3-2 U.S. NRC Standard Review Plan, NUREG-0800. 3-3 Shogan, R. P., Albertin, L., Yanichko, S. E., Lippencott, E. P., "Analysis of Capsule X from the Alabama Power Co.. Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program" Westinghouse Electric Corporation, WCAP-11563, Sept.1987. 3-4 Burr,s, N. L., Magee, R. D., Suggs, C. W., Jusino, B. J., "Residual Heat Removal System Autoclosure Interlock Ro'novs1 Report for the Joseph Farley Nuclear Plant Units 1 and 2, "Westinghouse Electric Corporation, WCAP-11746, to be published. me.m o 3-12

r. TABLE 3-1 LTOP EVENT TREE FOR BEST ESTIMATE CALCULATION FREQUENCY OPC TMP OPS WSO ALR OPA CATEGORY (per reactor year)

                                                                    • 1 OK 0.3
                                                      • 2 OK
  • 6.0E-03
                                                      • 3 OK
  • 0.90
  • 0.99 ****** 4 OK
                • 1.0 0.20 *
            • 5 OVPRES 4.6E-05 r
  • 0.01 ****** 6 OK
                • 1.0
            • 7 OVPRES 4.6E-07 8.7E-05*
  • 0.80
                                        • 8 OVPRES 1.8E-05 CATEGORY DESCRIPTION OK OVERPRESSURE EVENT IS MITJGATED OVPRES OVERPRESSURE SPIXE OCCURS TOTAL FREQUENCY OF SIGNIFICANT OVERPRESSURE EVENT IS 2.3 E-05 / CALENDAR YEAR meu io 3 13

o TABLE 3-2 LTOP EVENT TREE FOR CONSERVATIVE CALCULATION FREQUENCY OPC TMP OPS WSO ALR OPA CATEGORY (perreactoryear)

                                                                    • 1 OK 0.6
                                                      • 2 OK
  • 6.0E-03
                                                      • 3 OK
  • 0.90
  • 0.99 ****** 4 OK
                • 1.0 0.20 *
            • 5 OVPRES 9.4E-06
  • 0.01 ****** 6 OK
                • 1,0
            • 7 OVPRES 9.4E-08 8.7E-05*
  • 0.80
                                        • 8 OVPRES 3.7E-05 CATEGOR'l DESCRIPTION OK OVERPRESSURE EVENT IS MITIGATED OVPRES OVERPRESSURE SPIKE OCCURS TOTAL FREQUENCY OF SIGNIFICANT OVERPRESSURE EVENT IS 4.7 E-05 / CALENDAR YEAR i.

3 14

0 10 a 6 4 2 5 10-1 g s 5 I N 4 [ 2 E 5y 10-2 a : s 4 CORE MIDPLANE

  • TO VESSEL 2* -

CLOSURE HEAD l l I I '10-3 -300 -200 -100 0 100 200 300 400 DISTANCE FROM CORE MIDPLANE (cm) Figure 3-1. Relative Axial Variation of Fast (E > 1.0 MeV) Neutron Flux Within the Reactor Vessel - me 3-15

Nozzle to Shell Welds r(lowestpoint) gg 22.53 19-8948 36903 3 10-894 gw 45' 4 4= au o gg 5 144.0* 3 19-894A 86903 2 T q 20.1" gg,gg4 20-8948 86919 7 3

  • 45 am y

l m l 44.75" E i 86919 1 20-894A Figure 3-2. Indentification and Location of Welds in the Joseph Farley Unit 1 Reactor Vessel m:.mem ie 3-16

i. 4 4 l EVENT OTHER CATEGORIZATION SCHEMES I FREQUENCY PLAIC* NRC ANS I RANGE CONDITIONS RG I.48 RG 1.70 51.1 52.1 53.1 l W rwyearl CATEGOnlES 10 CFR ASME Code' Rev.2 (N18.3 (N212 IN213 W Condition Normel PC-1 Normel Normel Normel CondiLion pg y p ,A 1 l PC-1 Moderate Condition Anticipated Frequency II Plant ] gg.: gg g, i Occurrences I n!.,_:.; Condition PPC B g g g gg, 1,4 3 u Erner8ency PPC Condition c soa Limiting Condition 19d Accidents Faults IV Limiting Plant 104 Feelted PPC Condition PC8 D l 104 Not Considered l

  • Tea, e.n.h. hor t n m n.a im i

.t the Asus c.d l Figure 3-3. Event Categorization Relative to Event Frequency i

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== ::s. ii =: ap . u: -4:: = O O.05 0.10 0.15 0.2c 0.25 DISTANCE FROM SURFACE d) Figure 3-4. Evaluation Chart for Outlet Nozzle to Shell Weld I i X Inside Surface Surface Flaw X Longitudinal Flaw X Outside Surface X Embedded Flaw X Circumferential Flaw l l am.mem o 3-18 .}}

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