ML20040C075
| ML20040C075 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 01/30/1977 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML13319A635 | List:
|
| References | |
| FOIA-81-313 NUDOCS 8201270236 | |
| Download: ML20040C075 (79) | |
Text
{{#Wiki_filter:- ,o o STEAM GENERATOR IllSPECTI0llS
== SAN ONOFRE NUCLEAR GENERdTING STATION UNIT 1 JANUARY 30, 1977 "v' 8201270236 810925 PDR FOIA UDELL81-313 PDR Documenil
b i.. =*;t r (, ; r-"a ;*~s L, 9' p-tI section 1 INIRODUCTIO:4 A::D SU:' MARY 1-1 I 0 DISCUSSION OF RESULTS OF SMALL RADIUC Section 2 STEAM GENERATOR TUBE INSPECTION PROGRAM 2-1 e Small Radius Steam Generator Tube 2.1 2-1 Inspection Results Contribution of Results Toward the Accom-2.2 plishment of Principle Objectives of Integrated Inspection Program 2-5 Section 3 DISCUSSION OF SAFETY ASPECTS RELATIVE TO j LOWER SUPPORT PLATE FLOWSLOT DEFORMATION (HOURGLASSING) 3-1 4 J Section 4 DISCUSSION OF STEAM GEi1ERATOR TUBES AND ANTIVIBRATION BARS IN U-BEND AREA 4-1 i 4.1 Inspection Finding and Antivibration 4-1 { Bar Installation D Examination of, Removed Antivibration Bar 4.2 and Tube Samples 4-3 i 4.3 Safet Aspects of Antivibration Bar Installation 4-7 Section 5 DISCUSSION OF PROGRAM OF CORRECTIVE ACTIONS 5-1 5.1 Upper Support Plate Flow Slot Deformation (Hourglassing) 5-1 5.2 Lower Support Plat Flow Slot Deformation 5-2 (Hourglassing) 53 Steam Generator Tubes and Antivibration Bars in U-Bend Area 5-2 5-3 5.4 Conclusions A-1 l Appendix 1 l ._/ t 5
I h 9 L-e7 ry - r-t { J 5 T a b i r_ Tit 1; Pare 2.1 Results of Eddy Current Inspection: 2-8 I s 31 Steam Generator Shell Stress Evaluation 3-5 4 4.1 Summary San Onofre Unit 1 Tube Plugging Criteria 4-14 l l 4,2 Number of Tubes to be Plugged 4-15 I l l I I 4 e es. "i-11
T'rLE Fr C O!!T E! "'c (continued) 5' LIST GF r!F,"rrT Firure Title Pace 2.1 Steam Generator Sketch 2-9 c.2' Support Plate Sketch 2-10 2.3 Steam Generator "A" Inlet 400 KHZ (Reduced Gain) Through 4th Support; 400 KHZ (Reduced Gain) Through U-bend and Through Outlet Side 2-11 2.4 Steam Generator "A" Outlet 400 KHZ (Reduced Gain) Through 4th Support 2-12 2.5 Steam Generator "B" 400 KHZ (Reduced Gain) Through 4th Support 2-13 2.6 Steam Gene'rator "C" 400 KHZ'(Reduced Gain) I Through 4th Support' 2-14 -( ) 31 Strain Intensity, 10 mil Radial Dent Finite Element Simulation of Tube Support Plate Strains 3-6 32 Dented Tubes After Burst Testing 3-7 i 33 Steamline Break Dynamic Pressure Time History 3-8 34 Tube Vibration Test Apparatus 3-9 4.1 Steam Generators "A", "B" and "C" 400 KHZ 4-16 Inspection Around U-bend 4.2 Antivibration Bar Arrangement 4-17 43 Damage on Antivibration Bar-4-18 4.4 Surface at Unaffected Location of AVB, 100X 4-19 4.5 Surface at Damaged Location of AVB, 100X 4-18 J 4.6 Damaged Area on OD of Tube 4-19 /m(,). 47 Microstructure of Tubing at Damaged Region-4-20 4.8 Knoop Microhardness Traverse 4-21 iii b 5; ' -
.a..o, i ~ c -- : - r = e Title Pare y,- : - c I SCE Stca: Generator Tube Vibration 4,9 Analytical !:cdels for news 24, 36 and 38 (Support Case 1) 4-22 4 4.10 SCE Stean Generator Tube Vibration Analytical Models for Rows 24, 36 and 38 (Support Case 2) 4-23 4.11 27 Series Steam Generator Tube Row 48 Case 1 (Mode Shape No. 1) 4-24 4.12 27 Serics Steam Generator Tube Row 48 Case 2 (Mode Shape No. 1) 4-25 4.13 SCE Steam Generator Tube Vibration Analysis Results 4-26 1 w e 4 c-iv .gi.. ? Qu"" a.
i SECTION 1 ( -"7nnnornim:
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~~: On September 2, 1976, a' steam generator tube failure occurred 2. The Virginir. Electric and Power Company's Surry Unit No. c: failure at Surry Unit No. 2 appears to be due to a combination of intergranular stress corrosion and high stress in the small radius steam generator tubes resulting from upper tube support plate material movement at the center of the tube support plate which causes flow slot "hourglassing". This tube support plate material movement appears to be a result of stresses in the tube support plate as a result of tube " denting" which in turn appears to be the result of the expansion ~of precipitate and corrosion products r3 deposited in the tube support plate annuli where the-tubes pass U through the tube support plate. As a result of the Surry Unit No. 2 steam generator tube failure, Virginia Electric and Power Company, Florida Power and Light Company, Consolidated Edison Company and Southern California Edison Company, with the participation of the Westinghouse NiectricCorporation, formulated an integrated steam generator tube inspection program to assure the in'tegrity of the small radius steam generator tubes. The principle objectives of the integrated inspection program include: 1. Establishing whether or not the condition is confined to Surry Unit No. 2, / kT 1-1 'e
4 ? "eH r"4 - cr~ _Licn between flow slo; "hourglassin._" .{ ~ 2:icn of small radius tu cn a; [ 2- ,. 2. y - -"~d h-rclates to t".e conditions observed at Surry Unit No. 2, [ and ?. Defining and imnlementing a program that will assure continued cafe operation of the affected steam ( s f generators. s By letter dated November 2, 1976, the Southern California t Edison Company provided the NRC with its specific program to ) investigate the possible existence in San Onofre Unit 1 of con-1,4 ditions believed to have led to the steam generator tube failure et Surry Unit No. 2. That program included: (1) handhole entry inspections, (2) upper bundle entry inspections, and (3) eddy current probing inspections. Subsequent to submittal of the Novemb'er 2, 1976 letter, it was also determined that a three (3) inch access opening at the upper tube support plate would be drilled in Steam Generator "C" for observation and measurement of s the flow slots in the upper tube support plate. 1 In addition to accomplishing the program of inspections dis-oussed above to ascertain the condition of the steam generator upper tube support plates, the following conditions have been thoroughly investigated and evaluated, and where required, l cor-rective actions completed: ] 1. Support plate deformation and flow slot "hourglassing" ^>. in the bottom two (2) tube support plates in Steam Generators "A" and "C", only, and 1-2
Increases in tune wall degradation at the antivibration r--- cnc :ninning of the antivibration barn themselves due to excessive vibration damage in all three (3) steam generators. Thl: report presents the results of the program of inspec-and the details of the evaluation of the conditions
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discussed above. Section 2 discusses the results of the small radius steam generator tube inspection program, Section 3 dis-cusses the safety aspects relative to the lower support plate flow slot deformation ("hourglassing"), Section 4 discusses the steam generator tubes and antivibrations bars in the U-bend area and Section 5 discusses the program of corrective actions to assure continued integrity of the steam generators. b) It has been concluded that San Onofre Unit 1 operations can be resumed following the current refueling / maintenance outage without compromise to the safety or an increase in th'e risk associated with the operation of the steam generators. The basis for this conclusion is that (1) there is no evidence in the San Onofre Unit 1 steam generators of the conditions that resulted in the failure of a steam generator tube at Surry Unit No. 2, (2) ~ there_are no safety consequences with regard to ste'am genggator. lower tube support plate deformation and ligament cracking, and (3) further tube and antivibration bar degradation due to oxcessive vibration damage will be arrested by the addition of new cntivibration bars. V 1-3
t F SECTIO: 2 ~ DlJ3DiEI: 07 EE33L'.'O OF SMALL RADIUS I -:gryrm70g oROGRAt' --er .7^-
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2,5 Result of Small Radius Stear Generator Tube Inspections Lj San Oncfrc Nuclear Generating Station has three Westinghouse
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W S ries 27 steam generators. These are vertical shell and U-tube
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- i steam generators, forty-five (45) feet long and approximately fif-jjl teen (15) feet in diameter.
Each steam generator contains 3794 f 3 steam generator tubes mounted in the twenty-three (23) inch thick jij 1-tubesheet, and restrained from vibration by four (4) tube support plates spaced four (4) feet apart along the vertical run of the / ~ tubes and four (4) antivibration bars in the form of two (2) stacked V's located in the U-bend region of the tubes. These l I steam generators have been in commercial operation since January, 1968 for a total of 2,328 effective full power days. A sketch of a typical steam generator is shown in Figure 2.1.. During the fifth refueling outage, which started on September 30, 1976, an inspection program was' implemented to assure the integrity of small radius steam generator tubes as discussed in Section 1. The results of.that inspection program were as follows: 1. Handhole Entrv Inspections i ~ Routine eddy current inspection started on October 4, 1976. Because indications of defects were found in all three (3) steam generators, f.urther investigation of the 2-1
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- c-- go.2crators war conducted.
Photographs were t al:e n ~ 5' _.,.. _ ; ; 111 :r ee (3) stear generator by holding a ecr_ era inside eacn naadnoic witn the lens facing up. Each steam generator has two (2) handholes located I approximately twelve (12) inches above the tubesheet on opposite ends of the line of flowslots. The location of the handholes is shown in Figure 2.1 an1 a typical tube 1 support plate is shown in Figure 2.2 with the relative l position of the two (2) handholes identified. j l With the hand held camera, the two (2) flowslots nearest to the manhole in the first and second tube support l plates in all three (3) steam generators were photo-graphed. The_s_e., pho_t_o.gralbs_show deformation and. flow- . slot "hourglassing" in the bottom two (2) _ tube support plates in Steam Generators "A and "C", only (the safety _. aspects of thi.s condition are discussed in Section 3); there was no evidence of tube support plate deformation or flowslot "hourglassing" in Steam Generator "B". In order to photograph the flowslots in the upper two (2) tube support plates, the handhold entry technique discussed above was refined by use of a periscope apparatus With this apparatus, all the flowslots in the ,s upper two (2) tube support plates in Steam Generators "A" and "B" were photographed (photographs of the upper tube support plate flowslots in Steam Generator "C" were taken through the three'(3) inch access opening as 2-2
di_0ursed in iter h below). These photographs show no i ...;cm;._:n er fic. clot "hourglassing" in the upper two (2) tube support plates in Steam Generators "A" and "B". The photographs discussed abeve are included in Appendix 1. 2. Upner Bundle Entry Inspections An upper bundle entry was made into Steam Generator "A" to confirm the conclusions drawn from the photopraphs which were taken from the handholes. Using a boroscope, a videotape was made of the upper tube support plate flowslots. The videotape confirms that there is no deformation or flowslot "hourglassing" in the upper tube support plate in Steam Generator "A". D 3 Eddy Current Probing Inspections During December, 1976, 100 KHZ eddy current testing was ^ conducted in Rows 1 through 5 around the U-bend in all three (3) steam generators. Approximately 500 tubes were tested in each steam generator. This testing was designed specifically to detect-the presence of cracks at the apex of small radius steam generator tubes. No indications were found around the U-bend in any tubes tested. In addition to the 100 KHZ eddy current testing, a 400 KHZ eddy current reduced gain dent survey was conducted in each steam generator through -the upper tube support 2-3
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. l e n _1 u.- . t i : - .e tc e n n:^n :erc inCPe0 Led in eLOh - i stear generator. The results of the survey show that lj n d e r.t i r.r " i t c o r fir.c d to the first and second tube l Suppor; platen in 5 eam Genera; ors "A" and "C", and that there is no " denting" at the upper two (2) support i plates in these steam generators. For Steam Generator j "B", the results of the survey show no " denting" at any 1 of the tube support plates. The results of the eddy current probing inspections dis-cussed above are shown in Table 2.1. This Table also shows the results of the eddy current probing inspec- ~ ) tions conducted in the area of the antivibration bars which are discussed in Section 4.1. 4. Three (3) Inch Access Opening Inspections A three (3) inch access opening was drilled in Steam Generator "C" to allow observation and measurement of the flowslots in the upper support plate.. The access opening was drilled to facilitate future inspections and to allow physical measurement of the flowslot widths. { The opening was drilled through the shell'and wraper from the manway side of the steam generator just above 3 r i the upper tube support plate.'* Using a b'oroscope, a ~.g-videotape.was made of the upper tube gypport plate flowslots and the three (3)~ nearest flowslots in the } third tube support plate. The videotape shows no 2-4
deformation or flowslot "hourglassing" in~the upper two
- :,' tune suppert platec.
Ir addj; ion, to tne viaeotape, ph::cgraphs of the upper tube support plate flowslots were taken which also show no degradation or flowslot "hourglassing". The three (3) flowslots nearest the access opening were measured to determine their widths; these were the only flouslots accessible for measurements. Based on these measurements, the flowslots have suffered no deviations from the as manufactured condition, a measured width of 2-3/8 1/16 inches versus a manufactured. width of ~ 2-3/8 .120 inches. ~ The photographs discussed above are included in Appendix C' ), 2.2 Contribution of Results Toward the Accomplishment of Principle Objectives of Integrated Inspection Program Section 2.1 presented information to show the extent of tube " denting" and upper tube support plate flowslot "hourglassing" as I it exists at San Onofre Unit 1. This should be put in perspective j with the information that has been presented to the NRC staff dur-ing earlier conversation concerning other plants involved in_the i i integrated inspection program. On these other plants, the con-dition of the upper tube support plate and flowslot "hourglassing" ( have been discussed, mainly a's they affect the small radius steam i generator tubes. These effects have been quantified, analyzed ,a \\. 2-5
r-r --~-r . ::: cote: pl_r-- -.: al - ; -- r:- :- (-~ W-inspectier te monitor any kind of future degradation. =; m Very corrrenensive inspectient concucted at Scr Oncfre Unit 1 have shown that Steac Generator "B" nas no "centing"~ L[_ wnatsoever, W: and that the " denting" that exists in Stea= Generators "A" and "C" i:O H. w is confined to the lower tube support plates. Furthermore, the dssi w F::s " denting" which has occurred is modest by comparison with the g;; Tau N1 extent and degree of " denting" in the other plants where NWJ hh significant " denting" has occurred. San Onofre Unit 1 has utilized phosphate treatment in the secondary _ system as at cor-3d I roa-ion inhibitor, since commercial operation began in 1968. In addition, the inflow of impurities to the steam generaters from condenser inleakage has been significantly reduced as a result of retubing the condensers with ninety - ten (90 - 10) copper-nickel tubes and titanium tubes. The continued phophate treatment and improved condenser inleakage minimizes the environment conducive to tube " denting". Based on eddy current probing inspections, " denting" has increased at approximately 0.5 percent per year, and the current average " dent" depths of all tubes in the three (3) steam generators is three and one-half (3 1/2) mils. Additionally, the upper tube support plates, being key ones i insofar as small radius steam generator tube problems are concerned, have been shown consistently to have no defects or degradation. The inspection results show that the upper tube i l support plates have suffered no deviations from the as i
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\\ ( Based on the results of the inspection program discussed j above, San Onofre Unit 1 operations can be resured without compromise to the safety or an increase in the risk associated with the operation of the steam generators. In addition, since there is no degradation or " denting" in the upper tube support plates in any of the three (3) steam generators, neither a preventative tube plugging program nor an interim shutdown for reinspection prior to the inspection planned at the next refueling outage (see Section 5.1) are warranted based on the small radius steam generator tubes. D e 2-7
i. b .L ! I' ~ i p= er --t IN S P ECTIC'.'c RFSULTS M! Mi
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c." Tube Ir.rp ec te d P' Ster.r Gan. "A" Steam Gan. "P" Stenc Gan, "C" %I !! t/Cc'.c Hot / Cold Hot / Cold I { p,e % - ic7; 4?? ' m to 1st suppert 492/293 244/0 54/0 .j 4C; i:i::: around U-bend 870/0 1048/0 589/0 25 KHZ through 1st support 154/83 93/0 104/0 I 400 ):HZ through 4th support dent survey 724/385 0/0 287/0 b t' Dece P r 197f 400 KHZ around U-bend 905/15 831/0 1138/0 e 25 KHZ through 1st support 0/0 164/0 160/0 l 100 KHZ around U-bend 483/0 487/0 457/0 400 KHZ through 4th support dent survey 0/0 205/0 0/0 Antivibration Bar Indications Percent Indication Steam Gen. "A" Stean Gen. "B" Steam Gen. "C" < 20% 5 7' 6 21-29% 14 14 4 30-39% 47 7n 60 40-495 17 - 32 28 50-59% 2 13 16 60-69% 1 0 19 70-795 0 0 4 80-895 0 0 2 Total Tubes with AVB Indications 86 136 159 ' Indications at Individual AVB's Steam Gen. "A" Steam Gen. "B" Steam Gen. "C" 1. AVB i1 30 35 95 '2. AVB #2 25 47 F5 3 AVB f3 22 36 106 '4. AVB #4 16 30 90 5. Multiple 5 10 98 '100 KHZ Survey ] No indications were reported in the U-bend area of' those tubes tested in Stoem Generators "A", "B" or "C". l l l l u
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d _.. Z. ~ I e As discussed in Section 2, stear fencrator tube support plate ceformation and flowslot "hourglassing" at San Onofre Unit 1 is j 1 corfined to tne cottom two (2) tuce support platcs in Steam j a Generators "A" and "C", only. The existence of "hourclassing"_in j lhese tube support plates indicates the potential for cracking of tube support plate ligaments. The consequenec? of tube suppor_t plate deformation and ligament crachigg with regard to steam-I generator operational safety nave been evaluated. i 1 Finite element mathematical models were developed to quantify the relative strain levels throughout the tube support plate. The ptrforated region of the tube support plate is represented as an I aquivalent solid plate with reduced strength and stiffness properties. A Computer plotted contour map (Figure 3.1) displays strain intensities in the equivalent solid plate material. Strain intensities, as opposed to equivalent strains, are plotted because they provide a better measure of the state of strain in the plate with respect to its potential for tube deformation. As shown in Figure 3 1, the potential for tube deformation decreases with distance from the flowslots and outer wedges. Laboratory burst tests have confirmed the integrity of steam generator tubes under adverse. conditions of support plate dernrma-tion. Burst pressures of tubes with nominal wall thicknesses and cubjected to crushing loads (Figure 3 2) are not reduced below ) 3-1
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1. ...r: f : :.. : :. e r. i r - ry tne c u r r o u r.d i n g supper.- tinte cateric1. Evidence cf the restraint provided by the support I plate has been obtained through burst pressure tests of tubes with I reduced wall thic!:nesses in the area of tne support plate. The results show that burst pressures in tubes with up to fifty (50) l 1 percent wall thinning are essentially identical to those for un-i l I thinned and unsupported tubes. If the tUbc is not restrained by the support plate material, then internal pressure stresses could lead to tube burst only if the cracks exceed a critical size. Emperical expressions, derived from tests on unsupported and un-dented steam generator tubes with through wall axial cracks, reveal critical crack lengths in ' excess of 0 9 i'nches for E-} 0 750 x 0.055 inch tubes under accident condition (feed water line ^ break and steam line break) temperature and pressures. Critical crack lengths should not be significantly affected by support plate deformation. The most probable mechanism leading to cracking of support .1 plate ligaments results from the combined effects of tensile btressesandsupportplatecorrosion. Even in the presence of ligament cracking, the tubes and support plates are held in place by compressive forces. The safety aspects of potential support plate fragments fol-lowing a complete steamline break accident have been evaluated by enalysis and experiments. Hydrodynamic analyses (Figure 3 3) were ()I performed to quantify maximum fluid velocities and densities 3-2
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. r. e. ..:.; L ;r:lect_len, have verified tne rtructural in-tegrit;. Of the tubes. Tube fretting and abrasion tests were also l 6 carried out using the apparatus illustrated in Figure 3.4. Portulated accident conditions (design basis earthquake (0.67g), t steam line break, feedwater line break, loss-of-coolant accident) r were analyzed to obtain maximum values for tube vibration s amplitudes, contact forces and duration of loading. After tubes were forced to vibrate within a broken support plate annulus, they were burst pressure tested. Burst pressures of damaged tubes ware ( essentially unchanged. f Expansion of the support plate can impose loads on the steam I l generator shell through the channel segments located between the ) wrapper and the shell. Crush tests of tube support plates indi-cate maximum loads of 60,000 pounds at each of the six (6) channel locations. Since these loads are generated by the constraint of g ~ the surrounding shell, resulting stresses in the shell are classi-fled as secondary. Shell stresses derived from support plate expansion loads were combined with stresses from other loading l conditions, and the resultant stresses were compared with the American Society of Mechanical Engineering Code, Section III, allowable values. The stress levels, summarized in Table 3 1, indicate a significant margin available in all categories. Based on the evaluations discussed above, San Onofre Unit 1 operations can be resumed without compromise to the safety or an p increase 1n the risk associated with operation of the steam generators. In addition, since there are no safety consequences 3-3
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TALLE 31 STE/.M GENERATOR SHELL STFESS EV A LU ATIO!: Loading Loadings Stress Stress Stress Conditions Combined Category Level, Ksi Limit, Kai Design A+ C +E PL+ Pb 25. 1.5 Sm 40. = Normal & Upset B+ C + G(1) PL+ Pb+Q 25.(2) 3S =.80. m 84. Faulted A4 D+F PL+Pb 35. 3.15Sm = i Loadings: Design Pressure A = ~ a B = Operating Pressure. C = OBE Operating Moment D= DBE Overturning Moment E = OBE Channel Loads F = DBE Channel Loads G = Support Plate Expansion Loads Notes: 1. Greater of E and G included 2. Maximum range occurs at outside surface 3-5
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FIGURE I t-STEAMLINE B AK DYNAMIC PRESSURE TIME HISTORY I Dynamic pressure (Psi) t 1.30 - 1.20 - 1.10 - 1.00 - .90 - w .80 - E .70 - .60 - .50 - .4 0 - .30 - Location: top support plate Condition: no load .20 - w .10 - .05 - 0 ,T 0 1 2 3 4 5 6 7 Time (sec.) 10,154 20
TUBE VIBRATIONSEST APPARATUS .75" amplitude of I vibration for 10 seconds at 1725 RPM . integral ate A Vib. Broken tube {/ arm 7 support plate Uj %w gj M' N' N' ./ Y - 16.81 36 A 72 t 88 Removed i support 4 I_ Y Y sg ent /+ + + + + l l , o., s...,7
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ANTTVI:4-A T T O!' nA: ~ --- M.M L Ain 4.; Innnection Findines and Antivibration Bar Install.ation The progrcm of eddy current e::anination of the steam a generators included 400 KHZ inspections for tube wall thinning in the area of the antivibration bars (AVB) and reduced gain inspections for " denting" at the tube support plates. The results of the inspections for " denting" were discussed in Section 2.1. This Section will discuss the results of the inspections for tube wall thinning in the area of AVB's. Because indications of AVB related def'ects were found in all k three (3) steam generators, the number of tubes inspected around (} the U-bend.was increased to approximately 1900 tubes in each steam g en e ra. tor. AVB related tube defects in Steam Generator "A" occur mainly above row thirty-five (35); in Steam Generator "B", they ') occur as lcw as row twenty-one (21); and in Steam Generator "C", they are generally centralized in two (2) areas near the periphery of the tube bundle. Figure 4.1 shows a tubesheet map illustrating the area of each steam generator inspected. The results of the I eddy current inspections for tube wall thinning in the area of the .AVB's are shown in Table 2.1 of Section 2. I Because of the increase in the number of AVB related defects, and the accelerating rate of, tube wall thinning in the AVB area, three tubes and two AVB's were removed from Steam Generator "C" f and delivered to the Westinghouse Research and Development Center for examination to determine the mechanicm of wear. The results 4-1
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.,, a ;........... ...s o; a pa;:a:'.; a. . ; t L ' . g a 0 L a c r* at POiCOs Of contact. In order to arrest the accelerating tube wear in the U-bend aren. AVi's fa:r;;;tud fr:r ch r e r.e - p ; a t e d Inconel 600 with a square cross section instead of the original carbon steel bars with a circular crose section were installed between the existing AVB's. The new AVB's penetrate the bundle to a sufficient depth to provide additional support for all tubes between columns eighteen (18) through eighty-three (83) that have potential for excessive vibration damage. The new AVB's are located adjacent to the existing barn but sufficiently far away so that any new wear that may occur on the tube through contact with the new AVB's will _. ) not be superimposed onto the site of previous wear, ,l' Figure 4.2 shows an illustration of the new AVB arrangement. Section 4.3 discusses the safety aspects of the AVB modification. Initially an attempt was made to tighten the tube bundle by installing new AVB's with a nominal width of.267 .005 inches, as compared to the original design diameter of.250 .005 inches. licwe ver, the oversized bars could not be fitted between all columns of tubes. Therefore, AVB's with a nominal diameter of .250 inches were installed and the diametric clearance between the new AVB's and the tubes in the outer four (4) to six (6) rows were measured (these tubes were the only ones accessible far measurement). Whenever a large enough diametric 'alearance was N ceasured, the.250 inches AVB was replaced oy one with a width 4-2
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Inic 1; i: is tnc n:. inn; diametric clearance associated with the faoricaticn of AVB's for 51 Series Steam Generators. With this diarctric cicnrance, no unusual tube wall thinning has been tnese new units after several years of observec at AVB'c on ooservation. However, additional evaluations, as discussed in Section 4.3, indicate that the diametric clearance need not be limited to 0.019 inches in order to arrest further tube and AVB degradation due to excessive vibration damage. With the new chrome plated Inconel 600 AVB's with square cross-sections and the original as-built diametric clearance associated with the San Onofre Unit 1 Steam Generators which has not increased by the installation of the new AVB's, the tube wear rates will provide 1 _/ sufficient margin to the minimum uniform wall thickness required on the area of the AVB's. Accordingly, only about fifteen (15) percent of the AVB's with cross-sections of.267 inches were installed in an attempt to limit the diametric clearance to 0.019 inches; the remainder of the AVB's installed have a cross section of.250 inches and a diametric clearance equivalent to the original as-built clearance. 4.2 Examination of Removed Antivinration Bar and Tube Samples A portion of two (2) AVB's and sections of three (3) tubes which were in contact with the AVB's were removed from Steam Generator "C" and delivered to the Westinghouse Research and O 4-3
Lcycl;pacr.; Center for ona::.inction. Thc A'.73's were of carbon ... ni. T..Z5 ::ac:.cc act.ir.a1 diameter, and approximately three (3) feet leng; the tubes were identified as 348-061, R47-C61 and H46-C61. Tube R46-C61 had significant deterioration at the four (4) points of contact with the AVB surface and was chosen for de-l tailed exacinatior.. The results of the examinations were as i 1 follows: 1. AVB Examination a. Dimensions at affected locations: Micrometer measurements were taken at the various bar to tube intersections where metal removal had occurred. The minimum stock, remaining, at the worst location, was 0.076 inches, as shown in (]) Figure 4 3 In general, the maximum metal removal occurred near the periphery of the tube bundle. 4 I b. Visual examination of surface: The bar surface between the affacted areas was in relatively good condition. Some minor local attack was observed, but this possibly could have been caused by the solution used to decontaminate the bar at the plant. The surfaces at the affected locations were generally smooth and, in some cases, tright. -~ 2) 4-4
c. M~tallograpny at affected and undamaged locations: Met:11ccr:phic sectienc were taken thrcuch damaged and undamaged regions. The microstructure in both cases, shown in Figures 4.4 and 4.5, is normal. d. 141crohardness traverse: 1 A knoop 100 gr microhardness traverse was taken .l through a representative undamged location and through an adjacent damaged zone. Some' increase in hardness was noted at the demaged surface compared to that at the outside diameter of the undamaged region, 250 KHN to 180 KHN. This increased hard-ness diminishes with distanc_e from the affected surface. 2. Tube R46-C61 Examination a. Visual examination: Tube R465C61 exhibited the moct. aevere damage, at the AVB intersections, of the three (3) tubes; consecuently, this tube was selected for detailed examination. The other tubes exhibited only super-ficial damage. There was no obvious degradation of any of the tube sections with the exception, of j course, of the damage at the worst locations on l R46-C61. This area was electrolytically descaled to remove the adherent scale on the outside a diameter surface adjacent to the damaged 4-5 M "
_- i t r.. An the,n in Firure 4.6, the surface is ur.L;;aci:ed with the fine nolishing marl:s still ~- present frc: the production polishing operations performed in tube manufacture. b. Rcdiography: Short sections of tubing, containing the damaged areas, were removed and slit longitudinally, such that each half-section contained a damaged region. These areas were radiographed to determine if cracks (or other defects) were present in or near the damaged regions. No such defects vere observed. 4 c. Metallography:
- O Sections were taken through the damaged regions, both par,allel and transverse to the major axis of the elliptical sections.
These sections were then examined metallographically, both optically and with the scanning electron microscope (SEM). Figure 4.7 shows the etched structure at the 1 damaged region; the microstructure is normal with no evidence of cracking or corrosion attack at the damaged surface. The SEM examination tevealed a smooth surface with slight ripples; there was no evidence of corrosion attack. 4-6 e
d. EDAX-surface analysis: 1 & EDAM re:r Of a dent.ged region indicated only the presence of the major alloying elements of the tubing, nickel, chromium and iron, with lesser but typical concentrations'of titanium and phosphorus. e. Microhardness traverse: i Several knoop hardness traverses were taken through 4 a damaged region and compared with a traverse. i through an adjacent undamaged location. As shown in Figure 4.8, the material at and near the damaged region is harder than at the unaffected location. Based on the examination results discussed.above the damage i of.both the AVB and the tubing was caused by a wear mechanism due ]!C) to relative motion, probably impact, of the tubing and'AVB. This conclusion is supported by the increase in hardness observed at the entact surfaces and by the complete lack of any evidence of corrosive attack at the intersecting-locations. 43 Saf_c_tv Aspects of Antivibration Bar Installation ~ Section 4.1 discussed the results.of recent eddy current in-spections which indicated increases in tube wall degradation at 1 the AVB's in all three (3) steam generators. Thinning has alsb been observed on the round carbon steel AVB's, leading to increased clearances between tubes and the AVB's. The cause of-the increased tube thinning is the-result of excessive tube vibration and impact within increased clearances as discussed in Section 4.2. Hence, some tubes were essentially unsupported at AVB locations. 4-7
- rrftie tubc -!! stir- , hnniece th t 2 uie cecur in ^^ '"- ,uc.sr;;;rn, ;ne tuc (2} that are coct likely to cause danc-ing vibrations in conventional heat exchangers cre vortex rhedding and fluidelastic excitation. In the U-bend region, where the flow is two (2) phase, the vortex shedding mechanism is not expected to cause damaging vibration. Consequently, the tubes were evaluated for fluidelastic excitation. Finite element models were constructed to evaluate the effects of crossflow over tubes with different level radii and support conditions which simulated the existing and new support conditions. Tubes with bend radii corresponding to rows 24, 36 and 48 were modeled as shown in Figures 4.9 and 4.10. gS Support Case 1 represents a tub'e which is supported at each V AVB intersection. Thus, Support Case 1 is representative of the support conditions with the new AVB's. installed into the tube bundle. Support Case 2 depicts a tube which is unsupported at the bottom bar to simulate an excessive clearance between the tube and an existing AVB. Fundamental vibration mode shapes for each support case are illustrated in Figures 4.11 and 4.12. The results of the tube vibrator an'alyses, given in Figure 4.13, are generally consistent with eddy current data. They con-firm that tubes in rows thirty-two (32) through forty-eight (48) are unstable with respect to a fluidelastic vibration mechanism when they are not supported at the bottom AVB. Futhermore, all tubes are stable when supported at AVB's. 4-8
In additior, to the vibrational analyses on the tubes, the re: n-tisl for (1) excess vioration of tne AVB's the ceives, and (2) crc;tir.g loose parts should the already thinned AVB's fail at one (1) or two (2) points was evaluated. The likelihood of tube damage caused by AVB, as opposed to tube vibration, is extremely remote both for existing and new conditions. The presence of tube and AVB wear at the outermost tubes adjacent to the attachment of the AVB to the retainer ring confirms this. The addition of the new AVB's to the bundle does not significantly alter the tube spacing. Consequently, support conditions for the existing AVB's are not effected by the insertion of the new AVB's and excessive AVB vibration is not anticipated. Further thinning of the existing carbon steel AVB's will be arrested by the addition of the new Inconel 600 AVB's. In the case where an old AVB has O already failed at one or two points, resulting in a loose. piece, analys'is has shown that the normal operational flow forces will not lift the fragment. Therefore, impact of the fragment against tubes is not expected. Projectile impact tests simulating a worst case steam line break accident condition indicate the tube would not fail. Because of the increase in the number of AVB related defects and accelerating rate of tube wall thinning in the AVB area, some tubes in this area must be plugged. The tube plugging criteria (Table 4.1) is based on analysis, testing and operating history. As-a result of these evaluations, defect levels of fifty-three (53) percent at the upper tube support plate and sixty (60) O MI/ percent at the AVB locations and.other regions are acceptable. 49
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.r i r_. Uncer Cc Lined LOCA Plu 2SE Conditient, combined.titn design basis eartncuake (DBE) bending c resses derived from a seismic tire.ir: cry, analysis of tnc ::fified San Onofre Unit 1 stea: generators to a 0.67g ground :: tion earthquake. For this condition, the minimum uniform wall thickness required to meet American Society of Mechanical Engineering (ASME) Code, Section III, Faulted Conditions criteria for LOCA + DBE is 0,026 inches or forty-seven (47) percent of the nominal wall thickness. Analysis of the Faul'ted Condition loads (LOCA + DBE) employed-the use of the shape factor, 1.3, in the determination-of safe limits for 4 primary bending stresses (PL+P) in the tubes. This approach is { b I ) consistent with criteria defined in the ASME Code, Sect. ion III, and does not conflict with the proposed Regulatory Guide 1.121, Bases for Plugging Degra'ded PWR Steam Generator Tubes, August, t 1976. At AVB and tubesheet locations, where these bending stresses are less, the minimum wall thickness required is 0.020 inches or forty (40) percent. Tubes with wall thickness equal to 0.020 inches at AVB and tubesheet locations also meet ASME Code, 1 S ction III, limits for all other Faulted Condition loadings. i ) Furthermore, under normal operating pressures the primary membrane j l stresses do not exceed the yield stress of the tube material at I i t operating temperature, and the factor of safety against failure by l bursting exceeds three (3). ~ 4-10
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.10 : : ir ::1ur.: cigntesn (13) 1.*rcugr eighty-three (83) will be comparable to the wear rates experienced in stear generators operating with similar AVB designs (chrome plated Incenel 600 with square crcss-sections and 0.019 inches diametric clearance). After more than five (5) years of operation in steam generators with similar AVB designs, no unusual wear has occurred in tubes at AVB's. Initially, the diametric clearance between accessible tubes and AVB's was limited to 0.019 inches, based on Series 51 Steam Generators. However, after examining the eddy current data from 1967 through 1973, tube wear rates will be similar to, and certainly no worse than, those that occurred during early operation of San Onofre Unit 1 even assuming the original as-built diametric clearance associated with fabrication of the San Onofre Unit 1 Steam Generators. The original as-built diametric clearance has not been increased by the addition of the new AVB's. Another consideration examined in determining the acceptability of the original as-built diametric clearance involved the location of new wear that may occur on the tube i through contact with the new AVB's. Since the added AVB's will not be located precisely at locations of existing AVB wear, a new wear, if it does occur, will not be superimposed onto existing tube defects. If it is conservatively assumed that the tube wear rates will be similar to those that occurred during the earzy operation of the San Onofre Unit 1 steam generators (based upon eddy current data from 1967 through 1973), then a five (5) percent 4 11
_ ;... _ : r 111:v:n:: it rc:Uir-d : 7:r ri t h the oririnal - - e;.: :10.rr;r e. T." e fifty (50i pcreent pluggint mar.in of ten (10) percen: for tube wall r _ ;... a degradation between inspection shutdowns. in rows thirty Special consideration is given to the tubes (30) tnrough thirty-six (36), columns twelve (12) through seventeen (17) and eighty-four (84) through eighty-nine (89). These tubes are not supported by the new AVB's, and due to their large bend racii, the five (5) percent per year degradation allowance does not apply to these tubes. Tne maximum anticipated rate for these tubes is less than forty (40) percent per year based on previous operating history. Any of these tubes with i measured indications, even less than twenty (20) percent,.will be
- ,'N plugged.
This will provide a margin of sixty (60) percent for ! \\m) tube wall degradation between inspection shutdowns. With respect to the' upper tube support plates, there has never been any evidence of tube thinning at that location; and thereforo, no explicit allowance for degradation between inspections is deemed necessary. However, a fifty (50) percent plugging criteria will still provide a three (3) percent margin in ( any event. Similarly, a fifty (50) percent criteria for other defects would provide a ten (10) percent allowance for degradation between inspections which, based on past eddy current data, is more than sufficient. Table 4.2 shows the number of tubes to be plugged assuming 4()) the plugging criteria discussed above. 4-12
r.2 -Sirne and corriderttione discussed above, _gr_- ,use anc lE degradation due tc excescive vibration dcccge will bc arrested. Tnerefore, San Onofre Unit 1 <.perations can be resuaed without compromise to the safety or an increase in the In risk associated with the operation of the steam generc;ces. further corrective actions or an interim shutdown for
- addition, reinspection prior to the inspection planned at the next refueling (see Section 5 3) are not warrant.ed based on the tubes and
~ outage AVB's in the U-bend area. s , w M 4 5 9 4-13
t I 3;yur v s;. t: O ', r E.' j r, .,.,..- t .au a - tor-( l ,l l TOP AVE AVb ALL DEFECT LOCATIO!! PLATE (1) (2) OTHER Plugging Limit 50% 50% 100; 50% .i Min. h'all Thickness Required 47% 40% 40% 40% Margin 3% 10% 60% 10% NOTES: ) (1) Columns 18-23 (2) Rows 30-36, columns 12-17, 84-89 0 e l 4 h 1 I e-Re
S. G. S. G. S. G. A B C TOTAL > 50" t.VE DEFECT 3 13 41 57 > 50; TUBESHEET DEFECT 12 0 0 12 I!! TRI A!!GLEH 12 5 8 25 LEAKING 3 0 0 3 REMOVED FOR ANALYSIS t* 3* ~ 4 TOTAL 30 18 51' 99* ] PREVIOUSLY PLUGGED 21 11 49 81 GRAND TOTAL 51 29 100 180 TOTAL TUBES IN EACH STEAM GENERATOR - 3794 (*) THREE TUBES REMOVED FOR ANALYSIS - ONE CONTAINED AN AVB DEFECT > 505 (**) R0WS 30-36, COLUMNS 12-17 and 84-89 P O 4-15
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- Add V Bars Between Existing V Bars and Attach to Existing Retainer Ring NOTE:
The modification affects tubes in columns 18 through l 83 inclusive. l l l 6 ~ _. _ Figure 4.2 Antivibration Bar Arrangement 4-17
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Simple (Top V Bar) (Nodes 32, 36) i U-Eend !!c;;ign g y = 6.4 t/ft ~p D S1: rale (Bottom V Bar) U = 8.7 1p' 9N (Nodes 28, 40) s o ( 4 bN 'SdhN Simple (Top Support Plate) o o (N des 23, 45) Span = 45.25,, 3 y = 20 #/ft O o DC Sd)d DGDd Dd Simple (Nodes 19, 49) Span = 45.25" o o 3 y = 20 #/ft o ip 1 O O ~ DGDd)d DGDd DQ Simple (Nodes 15, 53) Span = 45.25" d 3 o o y = 50 #/ft o ~ o hQDQ DQDQD Simple (Nodes 11, 57) 4 e Span = 45.25" 3 Y = 50 #/ft Wrapper Opening 63/8f N U = 11.2 fps ,,p,,,,,,,,,,- Fixed (Tubesheet)(Modes 1, 67) Rows 48 36 24 24 36 48 Figure 4.9 SCE Steam Generator Tube Vibration Analytical Models for-Tube Rows 24, 36 and 38 (Support Case 1) 4-22
/ t-Simple (Top V Bar)(Nodes 32, 36) U-T,end Region [ C9 N. y = 6. l. I' / f t O ON \\ U = 8.7 fps \\ a e I Dh hd)d ~ Dd h simple (Top Support Plate) Spar. = 45.25" y (Nodes 23, 45) 3 o e y = 20 #/ft I o o DG Dd)G DG)G D'G Si=ple (Nodes 19, 49) y=20#/ft}" Span = 45.2 o o I o e DQDG)( DQDG DQ Simple (No' des 15, 53) 0 Span = 45.25"O y = 50 #/ft3 o o o o @ DGDG DGDG Simple (Nodes 11, 57) Span = 45.25" y - 50 #/ft3" Wrapper Opening +6 3/ + g + Fixed (Tubesheet) (Nodes 1, 67) + U = 11.1 fps em, <, u u < , <, <, u f,,q, Rows 48 36 24 24 36 48 Figure 4.10. SCE Steam Generator Tube Vibration Analytical' Models for Rows 24, 36 and 38 (Support Case 2). e O ~ 4-23 e- = y n.
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a W -H 0.0 i 10 20 30 40 50 TUBE BEND RADIUS, INCilES l l 1 10 24 36 48 TUBE Rn') NUMBER Pigure 4.13 SCE Stean Generator Tube Vib. ration Analvsis esults. 4-26
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a? C ", E r: rT : / r c-v;3 5.; Upper Tuot Su;oort Plate Flowslot Deforention /"Hourrlaccine") Section 2 discussed the results of the rmall radius steam generator tube inspcotion program conducted at San Onofre Unit 1. Based on that discussion, there is no evidence in the San Onofre Unit 1 steam generators of the conditions that resulted in failure of the steam generator tube at Surry Unit No. 2; therefore, neither a program of preventative tube plugging nor an interie shutdown for reinspection prior to the next refueling outage are warranted to assure integrity of small radius steam denerator tubes. During the next refueling outage, the Tcllowing inspec-tions will be performed to monitor the condition of the upper tube h support plates: 1. Photographing.flowslots in upper tube support plates in Steam Generators "A" and "B" thro' ugh handhole entry, 2. Eddy current probing inspections of all tubes in rows 1 through 5 around U-bend in Steam Generators "A", "B" and "C", arad 3 Photographing and measuring flowslots in upper tube support plate through three (3) inch access opening in Steam Generator "C". O 5-1
l 5.2 Lower ".~ur.e Support ?ltte Flowelct ...-...,s i 9 s c e t i er. }.;scunaec tne cafety Lspects relatite tc lower tube g suppcrt plate flow: lot deformation ("hourglassing"). _Based on that discussion, there are no safety consequences with regard to 1 4 steatn gener;ter tube support plate deformation and ligament crack-ing; therefore, neither a program of corrective actions nor 4 a ,a_dditional inspections are warranted to assure continued safe 1 operation of the steam generators. 53 Steam Generator Tubes and Antivibration Bars in U-bend ~ Area Section 4 discussed the program of corrective actions being implemented to assure that further tube-and AVB degradation due to 4 excessive vibration damage will be arrested. These corrective actions consisted of: l O 1. Installation o'r new AVB's of different size, material y and location, and 2. Preventative tube plugging for all affected tubes not supported by new AVB's, rows thirty (30) through thirty-six (36), columns twelve (12) through seventeen (17) and eighty-four (84) through eighty-nine (89), with measured indications, even less than twenty (20) percent. In addition to these corrective actions, eddy current probing inspections of all tubes in the area of the AVB's will be per- ) formed during the next refueling outage. The eddy current probing inspections will be performed on all tubes in rows twenty (20)
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