ML19275A052
| ML19275A052 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/22/1979 |
| From: | Counsil W, Switzer D NORTHEAST UTILITIES |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| TAC-11793, TAC-43142, NUDOCS 7908300642 | |
| Download: ML19275A052 (46) | |
Text
{{#Wiki_filter:. , . NORTHI!AST IFI'II.ITII!S '*
'd' P O Box 270 HARTFORD. CONNECTICUT 06101 4 (2'13) 666-6911 k L J August 22, 1979 Docket Ib. 50-336 Director of Nuclear Reactor Regulation Attn: Mr. R. Reid, Chief Operating Reactors Branch #4 U. S. Nuclear Regulatory Commission Washington, D. C. 20555
References:
(1) B. H. Grier letter to W. G. Counsil dated June 25, 1979, transmitting I6E Bulletin #79-13. (2) W. G. Counsil letter to R. Reid dated June 18, 1979. (3) W. G. Counsil letter to B. H. Grier dated July 13, 1979. Gentlemen: Fdllstone Nuclear Power Station, Unit No. 2 Feedwater System Piping In Reference (1), the NRC Staff summarized relevant industry experience regarding examination of feedwater piping welds, and requested action of Northeast Nuclear Energy Company (NNECO) to address the postulated increased likelihood of a feedwater line break. Due to an unscheduled cold shutdown which commenced on August 9,1979, NNECO initiated efforts to respond to Reference (1) by con-ducting radiographic examinations of the Millstone Unit No. 2 feedwater system piping welds. As discussed in a meeting in your offices in Bethesda on August 21, 1979, linear indications in the circumferential direction were detected on the heat-affected zones of weJds AC-G-1 AC-G-2 BC-C-1 BC-G-2 These welds are 2dentified on the isometric provided on Figure 1. The if.diographic testing results for these four welds are provided in Table 1. ihese welds are the safe-end to pipe and pipe to elbow welds in the horizontal run of piping connected to the steam generator feedwater nozzles. Once it was determined that these linear indications existed on the safe-end to pipe welds, NNECO expanded the scope of the radiographic examinations to include feedwater 2032 J03 7 9083 0 0Giilt .
piping welds up to the first piping support or snubber and high stress points in containment in accordance with the requirements of Reference (1). These examinations resulted in the detection of the two additional linear indications found at the pipe to elbow welds of both loops. A visual inspection of the feedwater system piping supports and snubbers in containment has also been completed. Operability and conformance to design has been satisfactorily completed in all cases. The nozzle to safe-end and safe-end to pipe welds of both steam generators are located within a 24" diameter pipe sleeve which extends approximately five feet through the steam generator shield wall (Figures 2 and 3). Location is such that removal of approximately two feet of concrete and reinforcing members around the sleeve to a depth of approximately six inches would be required for access to replace this pipe section. The details of the program will be provided in subsequent correspondence as discussed later in this letter. The closure weld would become the safe-end to nozzle weld (Figure 4), requiring substantial machining and stringent fabrication controls to prevent damage to the steam generator nozzle or shell. It is estimated that this repair would require four-to-six weeks to accomplish at this time. Another significant con-sideration involved in assessing these linear indications is the absence of identified corrective action to eliminate further crack initiation. Replacement with similar materials may result in indications of similar magnitudes in a rela-tively short period of time. The thermal fatigue analysis presented as Attachment 1 gives further credence to deferring replacement at this time. NNECO has obtained the services of the Westinghouse Electric Corporation to evaluate the radiographic examination results and to perform the ultrasonic examination of the af fected weids. The ultrasonic examinations were performed using 60' and 52" shear waves, and were referenced to cracked piping samples of known crack dimensions. The examintion of weld BC-G-1 confirmed the location and extent of the sharpest indication detected by radiographic techniques. The utilization of cracked samples as a basis for ultrasonic examination provides greater assurance of the reliability of the measurements. The maximum depth of the crack was 0.10 to 0.11 inches, as compared with the actual pipe thickness of 0.840 inches. The details of the ultrasonic examina-tion methods, procedures, and results are provided as Attachment 2. Table 2 is attached to demonstrate the margin which exists at the Millstone plant from allowable stress limits for the feedwater piping system. Note that the thermal stresses listed are for locations remote from the nozzle to safe-end juncture. Normal operating stresses at the nozzle to safe-end locations are calculated to be considerably below code allowable limits. Westinghouse experience on seven out of ten plants has shown indications of comparable depth and orientation to those presented herein. The probable mechanism for these indications is the same, independent of NSSS design, in that most PWR's have similar feedwater and auxiliary feedwater system designs resulting in comparable thermal cycling of the affected piping. The thermal 2032 004
= , fatigue analysis provided as Attachment 1 supports the - 'ention that the cracks will not propagate beyond approximately 100 mils in depth as a result of this rapid thermal cycling. The auxiliary feedwater system operating history of Millstone Unit No. 2 is also significant in light of the postulated relationship between linear indications and thermal cycling. The auxiliary feedwater system piping joins the main feedwater piping in the piping penetration room outside containment. The auxiliary feedwater pumps, one 100% capacity steam driven and two 50% capacity motor driven, take suction from the condensate storage tank (CST) . The typical temperature range of the CST water is from 50*F to 80*F. This fluid is not preheated before it enters the steam generator. The uses of the auxiliary feedwater system at Millstone Unit No. 2 are summarized as follows. During a normal startup, auxiliary feedwater is used from the time primary system temperature reaches 532 F until approximately 2% pover, at which time the main feed pumps are utilized. Typically, the period of time involved is eight hours. During a normal shutdown, one main feed pump remains in service until primary system temperature reaches approximately 425 F, at which time the auxiliary feed pumps are used. The exceptions to this guideline are evolutions involving minimal decay heat or unavailability of the condenser, when auxiliary feed pumps are utilized as soon as primary systcm temperature reachea 532 F. Auxiliary feedwater system flows vary directly with the heat loads, but a typical value is 150 gpm per loop. Cursory reviews of operating records indicate that approximatcly 115 reactor trips / shutdowns have occurred during the life of the plant. Roughly 80% of these involved a transition from main to auxiliary feedwater. It is estimated that, including pre-core and post-core hot functionals, 80 - 100 days of plant operation using auxiliary feedwater have occurred. The vast majority of these plant trips and uses of au::iliary feedwater have occurred from the summer of 1975 to the summer of 1976. It would certainly be expected that plant operation during the remainder of Cycle 3 would be typified by plant experiences during the past several years rather than the initial year of operation. For example, between the first and second refueling outages, the plant was never taken to a cold shutdown (Mode 5) condition. NNECO also has concluded that the potential for water hammer at Millstone Unit No. 2 has been adequately addressed. Since the first, and only, water hammer incident occurred, the plant design has been modified to include J tubes on the feedwater spargers. In addition, modifications have been completed on the feedwater piping system configuration to further minimize this potential. Tests were conducted on the modified system to verify the absence of water hammer. These efforts resulted in the issuance of License Amendment 32. This amendment, which remains in effect today, includes a license condition which precludes the addition of auxiliary feedwater for condit ions not bounded by the tests conducted. Additional details regarding these modifi-cations, as well as other relevant data on the feedwater system, were provided in Reference (2). In the unlikely event of a feedwater hammer occurring before the October 31 shutdown, NNECO will initiate a controlled shutdown to perform an ultrasonic examination of the most severe linear indication on the affected line. 2032 005
Based upon the above information, the following - points and conclusions regarding the indications are drawn. (1) The cracks found at Millstone Unit No. 2 are very similar to seven out of the ten Westinghouse plants examined. That is, the maximum indication Caund is approximately 100 mils. (2) the 100 mil cracks present are consistent with the results of a fatigue analysis based on high frequency temperature fluctuations. (3) This fatigue analysis also reveals that crack depth will be self-limited to 100 mils, which is consistent with those observed. (4) The implementation of on-line monitoring will provide detection of any unanticipated condition which could contribute to crack growth. (5) Fracture mechanics analysis of the feedwater lines shows the critical size of a through wall crack is greater than 26 inches; and for a part through continuous crack is greater than 70% of wall thickness. (6) Therefore, it is concluded that it is unreasonable to assume failure of the feedwater line by growth of flaws to critical size. (7) Furthermore, the indications observed comply with the ductile failure limits of ASME Section XI, lWE 3600. (8) The existing leak detection eq ipment has adequate sensitivity to identify a postulated leak in a timely fashion. Details of the instrumentation involved were provided in Reference (3). NNECO's review of the above material has concluded that Millstane Unit No. 2 is suit-ble for continued, safe operation at this time, without repair / replacement of th welds identified in Table 1, In order to fully respond to NRC Staff concerns in this matter, NNECO hereby commits to the following: (1) By October 1,1979, NNECO will docket a thorough weld repair program. The repair program will also address the removal of a section of the steam generator shield wall. (2) By October 31, 1979, NNECO will commence an approach to cold shutdown for the purposes of performing a second ultrasonic examination of the sharpest indication adjacent to each of the four affected welds. The re-examination will be conducted under similar conditions, i.e. , temperature, empty-pipe, etc. (3) Monitoring equipment will be installed on the safe-end to pipe weld on steam generator #2 prior to returning the unit to service. The details
. of this instrumentation package will be provided in a letter which will be docketed as soon as the package is developed.
2032 006
The Millstone Unit No. 2 Nuclear Review Board has also reviewed the acceptability of returning the unit to service at this time, and has concurred in the above noted determination. A copy of this letter will be forwarded to the Office of Inspection and Enforcement in fulfillment of the reporting requirements of Reference (1), in accordance with the understanding reached between our respective Staff s during the August 21, 1979 meeting, the absence of future docketed correspondence from the NRC on this subject will result in returning Millstone Unit No. 2 to service subsequent to completion of commitment (3) noted above.
' aur timely and favorable action in this matter is most appreciated.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY L (: br t w s: .( W. G. Counsil Vice President By: - D. C. Switzer President } Attachments L !
B D DOCKET No. 50-3}6 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 FEEDWATER SYSTEM PIPING 2032 008 AUGUST, 1979
TABLE 1 MILLSTONE UNIT NO. 2 STEAM GENERATOR #1 WELD # AC-G-1 (SAFE-END-TO-PIPE WELD) Circumferential Locatio n* Indication Dictance From Component (RT Film) Length Centerline of Weld Safe-End 57 . 8 cm 1. 2 cm Safe-End 0-2 3.7 cm 1.0 cm Safe-End 12 - 14 Intermittent 1.2 and 1.4 cm Pipe 13 . 9 cm 1.5 cm Pipe 13 - 14 1.4 cm 1. 5 cm Pipe 9 .7 cm 1.3 cm Safe-End 13 - 18 8.0 cm 1.3 cm Safe-End 16-1/2 - 18 4.0 cm 1. 0 cm Pipe 14 - 16 5.7 cm 1.4 cm Pipe 12-1/2 . 9 cm 1.4 cm Pipe 14 1.5 cm 1.0 cm Safe-End 45 - 48 Intermittent 1.3 cm Pipe 45 1/2 3.4 cm 1.4 cm Safe-End 45 - 47 4.4 cm 1.2 cm Pipe 44 - 45 1. 2 cm 1.3 cm Pipe 45 - 46 .8 cm 1. 3 cm Pipa 46 - 47 2.6 cm 1. 2 cm Safe-End 46 - 49 Intermittent 1.3 cm Safe-End 49 - 50 1.8 cm .7 cm 2032 009
e n TABLE 1 (Continued) MILLSTONE UNIT NO. 2 STEAM GENERATOR #2 WELD # BC-G-1 (SAFE-FND-TO-PIPE WELD) Circumferential Location
- Indication Distanco From Component (RT Film) Length Centerline of Weld Safe-End 57 - 3 12 cm 1 cm Pipe 57 - 0 2 cm 13 mm Pipe 1 1 cm 13 mm Pipe 1-2 2 cm 11 mm Pipe 11 - 13 3 cm 2 cm Safe-End 7-1/2 - 8 ^-5 mm 2.5 cm Safe-End 12 - 14 3.5 cm 14 mm Safe-End 45 - 48 8 cm 1 cm Pipe 43 - 46 8 cm 1.3 cm Pipe 46-1/2 1/2 3 cm 1.2 cm Safe-End 56-1/2 - 2 9 cm 1. 0 cm Safe-End 55 - 56 3 cm 1. 2 cm Pipe 57 - 0 1.5 cm 1.0 cm Safe-End 47 - 49 4.5 cm 1.3 cm Pipe 48 49 2.5 cm 1. 2 cm Pipe 57 -2 Intermittent . 9 cm Safe-End 46 - 51 Intermittent 1.2 cm
- X-ray film location markers at two-inch increments spanning 360 degrees, no7-3
) l
Table .' Continued MILLSTONE ITNIT NO. 2 STEAM GENERATOR #1 WELD # AC-G-2 (PIPE-To-ELBOW WELD) Indi:ations detected as follows: Marker Locations: 6 - 18 40 - 48 1:00 - 4:00 o' clock 8:00 - 10:00 o' clock 2032 011
Table 1 Continued MILLSTONE UNIT NO. 2 STEAM GENERATOR #2 WELD # BC-G-2 (PIPE-TO-ELBOW WELD) Indications detected as follows: Marker Locations: 0-2 12 - 18 40 - 48 2-4 12 - 16 41 - 47 46 - 52 12:00 o' clock 3:00 o' clock 9:00 o' clock 2032 012
TABLE 2 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 FEEDWATER PIPING SYSTEM STRESS SUM &iRY MAX. STRESS (l) (KSI) LOADING LOOP 1 LOOP 2 ALLOWABLE A THERMAL (2)(3) 16.6 19.5 22.5 B DEADWEIGHT 1.1 1.6 C PRESSURE 5.6 5.6 D SEISMIC 3.7 10.6 TOTAL (B+C+D) 10.3 17.8 18.0 (1) ANALYSIS PERFORMED BY BECHTEL. (2) MAX. OF HOT STANDBY AND NORN\L OPERATION. (3) INCLUDES SEISMIC ANCHOR MOTION. 2032 013
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ATTACID1ENT 1 THERMAL FATIGUE ANALYSIS OBSERVATIONS ORIGIN OF SMALL FLAWS BEHAVIOR OF FLAWS IN FUTURE SERVICE PIPING INTEGRITY EVALUATION 2032 018
MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 THERMAL FATIGUE ANALYSIS As a result of the Bulletin 79-?.3 inspections conducted in August, 1979, a number of small surface flaws were detected at or near the counterbore on the inside of the feedwater nozzle safe end-to-pipe welds (AC-G-1, BC-G-1) and the pipe to elbow welds (AC-G-2, BC-G-2). The welds are shown in Figure 1. The linear indications detected by radiography were subsequently mapped and characterized by ultrasonic inspections. The largest flaJs were confirmed to be approximately 100 mils in depth and oriented circumferentially in both feedwater loops. As a result of these findings, a piping integrity anelysis was conducted on the feedwater lines of this plant. The cracks found in these 1.ines were very similar in nature to those found in seven other }[ plants co date. Over the last several months, considerable ef fort has been devoted to understandin5; the causes and mechanisms involved in the cracking process. The cracks present in the Millstone feed lines, as well as in the other seven }[ plants, can be explained by the presence of temperature fluctuations in the water. Two scenarios are possible. (1) Rapid temperature cycles resulting from unstable mixing of hot and cold water during low flow, or from back leakage past the thermal sleeve. (2) Larger temperature fluctuations from initiation and control of cold feedwater flow. The crack initiation curve used is that of the ASME Code - this design curve has been shown in tests by GE/EPRI to actually be a very good estimate of an initiation curve for the feedwater environment. This curve should be further modified to account for the stress raises produced by the counterbore. For purposes of estimating the cycles necessary to initiate a crack at a given stress, or AT, level, the design arve was lowered by 25% to take account of surface effects (Figure 1). b 0 } ()
From Figure 1, it is seen that it is clearly possible to initiate cracks by either of the scenarios mentioned. This matches field experience, and is consistent with the GE/EPR1 research on their feedwater nozzle cracking. The most likely cause, in our judgment, is bigh cycle thermal fluctuations, and it is likely that the cracks initiated early in life. If the cracks initiated by a rapid cyclinb of the water temperature, and these cycles cuntinue, the cracks will not progress past the approximately 0.1 inch depth at which they are now. This is because rapid temperature fluctuations produce a significant range of cyclic stresses only within the first 0.1 inch of the wall, as can be seen in Figures 2 through 7. Shown in these figures are the temperature and stress profiles at a number of time steps, verifying the fact that the self-limiting depth does not change during a transient. Figures 4 and 6 show that the self-limiting depth is also independent of the magnitude of the temperature fluctuation. Thus, virtually any rapid ( nl Hz) fluctuation would be self-limiting and produce significant cyclic loads no greater thanN 0.1 inch into the wall, which correlates very closely with the observations on crack depths seen to date. The cracks can be expected to grow further in depth from further anticipated operational cycles of the plant, but this growth will be essentially insignifi-cant. To study how the cracks might grow, a series of fatig,e crack growth analyses were carried out. A crack growth rate law for carbon and low alloy steels in a water environment was used, which accounts for R ratio, and is shown in Figure 8. This law will be proposed as a replacement for the present Section XI crack growth law for steels in water environments, and is detailed in Reference (1). Results of the crack growth analyses are shown in Figures 9 and 10. Figure 10 shows the results of fluctuations in bending stress through the pipe wall, which would be most likely to result from thermal stress fluctuations. The magnitude of the stresses used can be seen from an example -- for a temperature fluctuation of 200*F, the resulting fluctuation in bending stress is 8 ksi. Figure 9 shows the results of a series of analyses considering fluctuations in membrane stresses in the pipe wall, as might result from bending of the pipe itself, or fluctuations in pressure. Another example is useful here - a pressure change of 600 psi produces a change Acm = 8 ksi. 00
i . It can be seen from these two figures that at least 1000 cycles are required to propagate these .100 cracks any further, regardless of the stress range considered. The largest stress range considered is larger than the stress range expected resulting from a startup-shutdown transient , of which only 200 are expected in 40 years. In considering Figures 9 and 10, two regimes of growth can be seen: (1) 10 3- 10 4cycles of very large fluctuations. (2) 105- 107 cycles of lesser stresses. These would have to occur at the rate of 300 - 30,000 per day to cause significant crack growth in the next one-year period. Piping Integrity Another important question to be answered is how the observed flaws compare with the size of flaws which could cause failure of the feedwater piping. To assess this subject, we must first define the mode of failure in this material, which is A106B carbon steel. Consideration of Charpy and Fracture Toughness data for this material indicates a transition to ductile behavior occurs at or below 0*F, far below the minimum operating temperature of the pipe. Therefore, piping fracture in the temperature range of interest will not be governed by linear elastic fracture -- it will be strictly plastic fracture, and is expected to be governed bv the plastic instability type failure. To evaluate whether unstable crack extension could occur before the limit load / plastic instability point, a Tearing Modulus analysis was carried out (References (2) and (3)). The maximum estimated Tearing Modulus (T) for conditions approaching limit load was equal to 5. A compilation of available data (Table 1) shows the smallest measuredmT at = 87, thus, giving a strong indication that the plastic instability made will govern failure. b2l
A plastic instability analysis was carried out using the design loads for Millstone Unit No. 2, including the design seismic loads. No water hammer leads were used because modifications have been completed to significantly reduce this potential. The results of the analysis are shown in Figure 11. Both part through and through wal; nritical flaw sizes were calculated, and the results show a through crack longet -han 26 inches is needed to cause failure, and a part through crack all the way around the pipe and 70% through the wall is required. Clearly, there is significant margin against failure. Furthermore, even if the cracks did propagate to a significant. extent, leakage would result before the crack propagated sufficient to result in a pipe rupture. This can be seen from the present morphology, which shows the cracks concentrated near the top of the pipe. This leak before break concept has also been confirmed by field experience with feedwater piping cracks. If the cracks grow, they do so only at a localized region around the pipe circumference. 2032 ')?2
TABLE 1
SUMMARY
OF TOUGHNESS DATA dJ cy ou U o JIC da T Source : Gudas-USNRDC (Sulfur - 017 .029)(Ref. 4) A106C RT 49.5 81 65.3 1380 20500 144 A516Cr70-RT 52.0 81 66.5 890 15000 101.7
-300F 40.0 69 54.5 664,850 12000 125.6 11500 Source : W Unpublished Data (Sulfur .004 - 0.010) all at Room Temperature 516cr70-RT 49 75.1 62.1 3030 59320 461 52 74 63 4400 29930 226.2 SA350LF2 Forging RW 81.3 66.7 4180 8528 57.51 Source: W WCAP 9499 (Ref. 5) (S = 0.030 .037) 516Gr70-RT 46.7 71 58.9 790 10745 93.07 440F 31.3 61 46.2 375 6082 85.4 RT 45.5 71 58.3 995 440F 32.7 61.2 46.9 995 16801 229.1 2032 023
REFERENCES (1) Bamford , W. H. , " Application of Corrosion Fatigue Crack Growth Rate Data to Integrity Analyses of Nuclear Reactor Vessels", ASME Trans Journal of Materials Technology, July,1979. (2) Tada , H. , Paris , P. , Gomez , M. and Gamble , R. , "A Stability Analysis of Circumferential Cracks for Reactor Piping Systems", presented at Twelf th National Symposium on Fracture Mechanics, St. Louis, MO, May, 1979. To be published as ASTM STP. (3) Bamford, W. H. , "An Assessment of the Impact of Thermal Aging on Integrity of Primary Coolant Piping", WCAP, January, 1979. (4) Palusamy, S., et. al, " Comanche Peak Steam and Electric Station Structural Integrity Evaluation of Mainsteam and Feedwater Super-pipes", Westinghouse Electric Corporation, WCAP 9499, Fby,1979. (5) Gudas, J. , personal communication, Fby,1979. 2032 024
FIGURE CAPTIONS FIGURE 1 . . . . . Initiation curve for fatigue crocks. Stress (or AT) vs. cycles. FIGURE 2 . . . . . Temperature profile at 15 seconds and 16 seconds. AT/sec = 50*F. Temperature vs. distance f rom ID wall of pipe. FIGURE 3 . . . . . Thermal stress profile at 15 seconds and 16 seconds. Stress vs. distance from ID wall of pipe. Note that Ao = 0.1". FIGURE 4 . . . . . Temperature profile at 17 seconds and 18 seconds. AT/sec = 200 F. Temperature vs. distance from ID wall of pipe. FIGURE 5 . . . . . Thermal stress profile at 18 seconds and 19 seconds. Stress vs. distance from ID wall of pipe. FIGURE 6 . . . . . Temperature profile at 57 seconds and 58 seconds. AT/sec = 200 F. Temperature vs. distance from ID wall of pipe. FIGURE 7 . . . . . Thermal stress profile at 57 seconds and 58 seconds. Stress va. distance from ID wall of pipe. FIGURE 8 . . . . . Crack growth rate as a function of cyclic stress intensity. FIGURE 9 . . . . . Crack depth as a function of number of cycles for various cyclic membrane stresses. ainitial = 0.1". FIGURE 10 . . . . . Crack depth as a function of number of cycles for v..rious cyclic bending stresses, ainitial = 0.1". FIGURE 11 . . . .. Unstable crack depth as a function of bending moment for two initial crack configurations. 2 l)s-) !,) _J
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ATTACIMENT 2 ULTRASONIC EXAMINATION MET 110DS , PROCEDURES , AND RESULTS 2032 037
Page 1 Rev. O U?,Th iS.;NIC TisI PROCEDURE F0F MAPPING ? HE
- JB 4PHIC INDICATIONS IN THE FEEDWATER 10ZZLF .mfE-Eh-TO PIPING JOINT, MILLSTONE UNIT 'L . 2 "9WF" PLA NT 1.0 SCOPE This proceduce describes the manual ultrasonic test prescribed for determining the location and length of the radiographic indications detected in the base metals adjacent to the feedwater nozzle safe-end to piping welds in the two loop Millstone Unit No. 2 power plant. The procedures necessary for estimating the depths of the radiographic indications compared to Elox notches are also included.
2.0 GENERAL REQUIREMENTS 2.1 Ultrasonic flaw dete~.'.on instruments shall be of the pulse echo type A-Scan presentation and shall be qualified to the requirements of Westinghouse procedure NSD-ISI-10 (attached) at the beginning of each period of extended use. The qualification may, under documented conditions, be considered valid for a period not to exceed three months. 2.2 Operators performing examinations to this procedure will be qualified to Level II in accordance with SNT-TC-1A, its supplement and appendices as applicable for ultrasonic testing. 2.3 The search unit to be used in the testing is a nominally 60* shear wave unit containing a 2.25 MHz,1/2 inch diameter transducer. The search unit will be provided (Westinghouse property) by Westinghouse for this testing program. 2.4 A suitable liquid or semi-liquid couplant, such as water or glycerine, shall be applied to the test surface. 2.5 The test surfaces shall be free of dirt, loose scale, machining or grinding particles, weld spatter, or other foreign matter that would impair the free movement of the search unit or affect the examination. 2032 033
, a Page 2 Rev. 0 3.0 REFERENCE BLOCKS 3.1 An IIu Elock shall be used to block out the basic calibration and to check the operating characteristics of the test system, if necessary. 3.2 A 10-inch x 5-inch x approximately 1-inch thick section of SA508 Class 2 material containing two Elox notches, one notch - 1/2 inch long x 1/16 inch deep, the other notch - 1/2 inch long x 1/8 inch deep. The block is identified by the stamping -- TVA 7 -- on one edge. 4.0 CALIBRATION 4.1 Prior to conducting the mapping examination, the complete test system (instrument , cables, the specified 60* shear wave search unit , couplant , etc.) shall be calibrated on the applicab2e block for the mapping operation. Once calibration has been established, any change to any part of the system will require at least verification of the calibration. 4.2 Sweep calibration. 4.2.1 Place the 60* search unit, properly coupled, on the 1.0 inch thick portion of the 508 Class 2 block (Paragraph 3.2) and set, by adjusting sweep delay and range controls, the maximized 1/2 vee path response from the corner of the block at increment 5 along the sweep (1/2 sweep length, position 5, marker 5, etc.) and the full vee path response at increment 10 of the sweep calibration. 4.2.2 Place 60* search unit on IIW Block and maximize response, by search unit movement, from 1/8 inch diameter side drilled hole (sdh). Adjust gain until response is 80% full screen height (FSH) amplitude. Decrease gain by 6 db to check vertical linearity and attenuator or gain control linearity -- response amplitude should be 40% FSH 5%. Determine gain control setting to achieve 1/2 amplitude response. 2032 03?
Page 3 Rev. 0 4.2.3 With search unit on IIW Block positioned for maximum response from the 1/8 inch diameter sdh, adjust gain to obtain a 15-20% FSH response. 4.2.4 Place search unit, properly coupled, on surface of TVA 7 (Paragraph 3.2) reference block opposite that containing the Elox notches. Maximize 1/2 Vee path response from the 1/8 inch deep notch -- will occur at position or increment 5 on sweep. Adjust gain, if necessary, to achieve an 80% FSH response. Mark this peak amplitude point on screen of cathode ray tube (CRT) . Nithout adjusting controls, maximize response from 1/16 inch deep notch (also at position 5 on Sweep) and mark peak amplitude (A 40% FSH) point on CRT. 4.2.5 At position 3.75 on Sweep, mark CRT screen at a point 1-1/2 times the peak amplitude response obtained f rom the 1/16 inch deep notch. This is the distance correction developed on a 1/16 inch deep notch in 3/4 inch thick piping material. Draw a line (DAC) through the two points based on the 1/16 inch notch respcases. This 1/16 inch deep notch DAC shall be consilared the p:1 mary reference. 5.0 SCANNING 5.1 Determine the centerline of safe-end to piping weld. Set or mark a reference p" Int at a set distance (7 inches) from the safe-end to pipe weld centerline on the nozzle. Mark or set a circumferential reference line at the set distance reference extending from the 9 o' clock position through the 12 o' clock position (top of pipe) to the 3 o' clock position. Mark off 2-inch increments on the reference line starting at 12 o' clock (0 zero) in both the clockwise and counterclockwise directions. 5.2 Scan with the search unit on the safe-end, sound beam directed toward the safe-end to pipe weld. Move the search unit to and from the weld on 1/4 inch increments around the pipe from the 9 o' clock position through 12 o' clock to 3 o' clock. 032 0/0
Page 4 Rev. 0 6.0 RECORDING 6.1 Record the peak amplitude of reflectors exceeding 1/2 the pricury reference amplitude. 6.2 Record the search unit distance from distance reference line drawn on nozzle. 6.3 Record 1/2 amplitude end points of reflectors by moving search unit circumferential1y from peak amplitude point. 6.4 Record the test metal distance (tmd) to the reflector at the peak amplitude point. 2032 De1
- 1 Page 5 2ev. 0 WELD # BC-C-1 AND AC-6-1 yJEAK AMPLITUDE AXIAL DIST. CIRC. LOCATION 1/2 AMPL. CW 1/2 AMPL. CCW TMD dLJ. J . , c.
ULTRASONIC EXAMINATIONS STEAM GENERATOR NO.1 PIPE TO SAFE-END WELD WELD # AC-G-1 Peak Axial Cire. 1/2 Ampl. 1/2 Ampl. Amplitude Dist. Location CW CCW TMD 54% FSH 5 14.5 in. 15 in. 12.5 in. 4 40% FSH 4-15/16 13.25 in. 15 in. 12.5 in. 4 50% FSH 4-3/4 46 in. 44 in. 47 in. 4 Signal Wa lk (TYP. ALL) 3.75 to 4.5 Major Divisions Thickness Measurements Pipe .760 in. Weld (Crown) .800 in. Safe-End 1.00 in. Measured 50% Points Dis tance Leading Trailing 14.5 in. + 1/16 - 1/8 13.25 in. + 1/ 8 - 1/8 46.0 in. + 3/16 - 1/8 Transducer : .5 in diameter, 2.25 MHz, 0 51 Scope : KBI U SM-2 Gain : 34 DBG Refernce Point: 7 in, from weld centerline 6-5/8 in, from crown edge 917^ i LJJ[ I h "n J
ULTRASONIC EXAMINATIONS (Continued) STEAM GENERATOR NO. 1 PIPE TO SAFE-END WELD WELD # AC-G-1 To read indication, increased sensitivity by 4 db 44 in to 47 in. 80 " SWP 4.0 - 50.5% FSH = 92% DAC = .058 in, deep
.8 12.5 in, to 17 in.
N SWP 4 . 0 - 7,
= 47.3 FSH = 86% DAC = .054 in, deep a t SWP 4. 0 DAC = 55% FSH O b 0 b b;-
, a ULTRASONIC EXAMINATIONS (Continued) STEAM GENERATOR #2 PIPE 'ID SAFE-END WELD WELD # BC-G-1 Peak Axial Circ. 1/2 Ampl. 1/2 Amp 1. Amplitude Dist. Location CW CCW TMD 95% FSH 5-7/ S in. O in. 2-1/4 in. 55-1/2 in. 4 85% FSH 4-7/a in. 46-1/2 in. 48-1/2 in. 43 in. 4.2 No amplitude above 25% FSH was observed from 2-1/2 in. to 15 in. Signal Walk (TYP. ALL) 3.75 to 4.5 Major Divisions. Thickness Measurements Pipe .760 in. Crown .800 in. Edge of crown on Safe-End .840 .880 in. Safe-End .960 in. Measured 50% DAC Points Distance Leading Trailing 56-1/2 in. - 1/8 + 1/4 0 in. - 1/8 + 1/4 2-1/4 in. - 1/8 + 1/4 44 in. - 1/8 + 1/4 46 in. - 1/8 + 1/4 47 in. - 1/8 + 1/4 1/16 in. notch - 3/16 + 1/4 1/8 in. notch - 3/16 + 1/4 Transducer : .5 in. diameter, 2.25 MHz , 0 51* Scope : KBI USM-2 Gain : 36 DBG Reference Point: 7 in, from weld centerline 6-5/8 in. from crown edge 032 0,15
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