ML20086C374
| ML20086C374 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 01/31/1973 |
| From: | Goldsmith E, Lehr R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| References | |
| NSD-E-NLB-672-R01, NSD-E-NLB-672-R1, NSD-NLB-672, NUDOCS 8311230221 | |
| Download: ML20086C374 (63) | |
Text
{{#Wiki_filter:_ -_- - 1 1[p) If &L a/o/zs H. B. ROBINSON STEAM GENERATOR TUBE LEAKAGE INVESTIGATION Revision 1 January, 1973 Edited by: Operating Plants Service Nuclear Service Division WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems x P. O. Box 355 Pittsburgh, Pennsylvania 15230 APPROVED: - 4 APPROVED: R. . Lehr, Manager E. A. Goldsmith Opera'ngPlantsfervice Engineering Manager
,l/NuclearServiceDepartment RETURM TO DIRECTORAT~. Y REGUu.J'j;l 0 5gATIONS 2 \
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- 8311230221 730131 PDR ADOCK 05000261 P PDR NSD-E-NLB-672
^
TABLE OF CONTENTS
- 1. Conclusions and Recommendations >
- 2. Description and Operating Characteristics of Robinson Steam Generators 2.1 Plant Characteristics and Operating History 2.2 Steam Generator Operating Conditions 2.3 Steam Generator Water Chemistry
- 3. Examination of Robinson Steam Generators 3.1 Eddy Current Examination 3.2 Removal of Tube Specimens 3.3 Metallurgical Analysis 3.4 Microanalysis of Tube-Wall Defects 3.5 Analysis of Sludge and Deposits
- 4. Probable Cause of Tube Failure
- 5. Corrective Measures to Prevent Reoccurences of Tube Leakage 5.1 Water Chemistry Control 5.2 Improvement in Circulation 9
l 1
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- 1. CONCLUSIONS AND RECOMMENDATIONS Detailed examination of the tube samples removed from the Robinson steam generators revealed an intergranular attack originating on the OD surface of the tubes. This attack was most probably a form of caustic stress corrosion.
A review of the generator secondary side operating history identified several factors which may have enhanced the environment contributing to the intergranular attack on the tubes. These include: a) reduced generator circulation ratio b) operation with condenser in-leakage without concurrent generator blowdown c) operation with frequent out-of-specification chemistry conditions d) phosphate concentration (10 ppm) which was inadequate in the presence of a), b) and c) above. The correction action that was implemented included: a) plugging of all generator tubes which had positive eddy current indications b) return of the steam generator circulation ratio to original design conditions c) increase the minimum generator phosphate level to 20 ppm d) continuous feed of phosphate, balanced by continuous blowdown The unit was returned to service on June 9,1972 and has been in operatior since that date. The primary to secondary leak rate since these repairs were made has been at the limit of detectability (0-6 gpd). This is the most positive assurance that the action taken to correct further degradation of the tubesb has been effective. 1
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- 2. DESCRIPTION AND OPERATING CHARACTERISTICS OF ROBINSON STEAM GENERATORS 2.1 Plant Characteristics and Operating History The H. B. Robinson Unit 2 station is a 2200 Mwth PWR (initial rating).
lne main coolant system contains three loops. Cooling water is pro-vided by Lake Robir, son, constructed specifically for this purpose. Figures 2.1-3 contain the power and temperature history for the Robin-son plant since initial startup. Highlights of plant operation are indicated on the Figures. Data are included up to the shutdown for steam generator repair in May,1972. During initial power escalation the primary system T program was ave reduced to permit operation with lower steam pressure. Periods of operation between February and September 1971 were with this reduced program. Full power operating data indicate a primary system T hot and T cold f approximately 590*F and 535 F respectively, with a steam pressure of 770 psia (yielding a steam temperature of 513*F assuming saturation conditions). In late September 1971, the RCS T ave program was again revised upward to the system conditions in present use. These are a main coolant T f Ol"F, T hot cold f 545 F, and steam pressure of 840 psia (Tsteam
- 524 F assuming saturation). These items have not changed significantly since tube plugging and steam generator modification in May 1972.
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2.2 S_ team Generator Operating Conditions A comparison of thermal and hydraulic conditions in operating Westing-house plants with inconel steam generators is made in Table 2.2-1. From the table it can be seen that, except for items related to the lower CPL circulation ratio, steam generator characteristics are similar to other operating plants. During the tube plugging operation in May,1972, the orfice plates in the CPL generators were raised to increase the circulation ratio,. The thermal and hydraulic characteristics and the calculated pressure stresses for CPL before and after the modification are shown in Table 2.2-2. Operating data from RGE is included for comparison. The calculations show the stresses to be within allowable code limits. Also, comparison with Inconel yield strength at operating temperature indicates the stress levels to be well below the yield strength of the materials, f
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Table 2.2-1 Thermal and Hydraulic Comparisons of ,0perating Plants o Plant UEM SCE CYW lK'K NBK RGE CPL WIS Units Date 12/28/70 8/22/69 3/26/69 'F/15/71 6/6/72 3/10/72 1/24/72 6/10/72 Mo/ Day /Yr Series 24 27 27 33 33 44 44 44 - Steam Pressure 676 826 651 752 834 786 838 854 psia Primary Inlet Temp 575.8 589.7 576.3 5S6.6 594.7 593.2 601.4 595.6 F Primary Outlet Temp 530.0 543.5 525.0 543.6 543.4 544.0 545.2 541.6 F Saturation Temp 499.2 521.9 495.1 517.1 523.0 516.0 523.5 525.8 *F Bundle Inlet Temp 450.0 497.2 476.9 489.4 494.7 483.9 480.5 498.7 F Boiling Inception Temp 486.6 516.8 487.2 E03.4 514.7 506.5 510.0 517.7 *F Tube Wall Temp 508.7 527.9 502.7 523.9 529.1 527.9 535.0 531.8 *F (Boiling) Tube Wall Temp 505.1 523.4 511.2 530.2 533.1 530.7 537.4 535.4 *F (Conv::ntion) Bundle Inlet Velocity 1.26 2.80 1.48 1.71 1.67 1.37 1.01 1.62 fps Crossflow Velocity 2.99 6.74 3.98 4'49. 4.38 3.61 2.66 4.27 fps (First Tube) Parallel Flow Velocity 0.58 i.27 1.07 1.20 1.17 1.07 0.80 1.26 fps Convective Heat Flux 1.13 0.96 0.92 l.10 1.02 0.99 1.02 0.96 5 10 8tu/hr-ft Boiling Heat Flux 1.07 0.89 1.04 5
- 1. 21 1.08 1.04 1.05 1.02 10 Btu /hr-ft 2 Height to Boiling 1.20 1.37 0.63 1.42 1.58 1.59 1.49 1.64 ft Inception 8
Table 2.2-2 Comparison of Heat Flux, Hydraulic Parameters, and Thermal and Pressure Strasses of,0perating Units Plant H.B. Rcbinson No. 2 R.E. Ginna No. 1 Reduced CR Increased CR Date 1/24/72 3/10/72 Units Primary Side Conditions Inlet Temperature 601.4 601.3 593.2 F Outlet TeTperature 545.6 545.0 544.0 *F Operating Pressure 2250 2250 2250 PSIA Secondary Site Conditions Feedwater Temperature 432.5 435.0 404.0 *F Steam Pressure 838 845 786 PSIA Steam Flo.. Rate 3.18 3.20 2.66 106 lb/hr Bundle Inlet Temperature 480.5 497.8 483.9 F Saturation Temperature. 523.5 524.5 516.2 *F Bundle Inlet Velocity 1.01 1.61 1.37 ft/sec First Tube Crossflow Velocity
- 2.66 4.26 3.61 ft/sec Heat Transfer Conditions
, local Boiling Heat Transfer Coefficient 1358 1358 1347 Btu /(hr-ft2, 7)
Local Boiling Heat Flux 105,000 104,000 104,000 2 Btu /(hr-ft ) local Con.iective Heat Transfer Coefficient 845 2 944 910 Btu /(hr-ft_.7) local Convective Heat Flux 102,000 98,000 2 99,000 Btu /(hr-ft ) Fouling Factor 0.00005 0.00005 0.00005 2 (hr-ft *F)/ Btu 9
Table 2.2-2 (con't) Flant H.B. Scbinson No. 2 R.E. Ginna No. 1 Reduced CF Increased CR Date 1/24/72 3/10/72 Units Tube Size 7/8 .050 7/8 .050 7/8 .050 inches Frimary Pressure 2250 2250 2250 psia Secondary Fressure 83E 845 786 psia Tube Pressure Stress 9343 9211 9769 psi Code Allowa;1e Stress 23,30C 23,300 23,300 psi Primary Inlet Temp 601.4 601.3 593.2 F Saturating Temp. 523.5 524.5 516.2 F Cundle Inlot Temp. 480.8 497.8 483.9 F Heat Fit x 3oili1g Convet tion 105,000 104,000 104,000 Btu /hr-ft 2 102,000 98,000 99,000 Btu /hr-ft2 Tube Resistivity 4.EE 4.58 4.60 10 4 hr-f t2- F/Ett Outside Film Resistivity Boiling 0.58 0.59 0.63 10 4h r-f t2- F/Btt Convectien 5.07 3.83 4.21 10-4h r-ft 'F/ Btu Tube Terperature Drop Boiling . 48.4 47.7 47.7 F Convection 46.8 44.8 45.7 F Tube Outside Temperature Boiling 535.0 535.9 527.9 F Convection 537.4 540.1 530.7 F Tube Thermai Stress Boiling 79E6 7851 7851 psi Convecticn nm 7703 7374 7522 psi Tube Total Stress Boiling 17,903 17,062 17,620 psi Convecticn 17,046 16,585 17,291 psi Code Allowable Stress 69,9C0 69,900 900 psi Inconal Yield Strength 27,900 27,900 69,900 27, psi 10
2.3 Steam Generator Water Chemistry The history of the H. B. Robinson #2 steam generator water chemistry during 1971 and 1972, both prior to and following the shutdown for repairs in May, 1972, is given in Figures 2.3-1 to 2.3-6. These plots present pH, phosphate, and sodium-to-phosphate ratios for each of the steam generators. During the period preceding the repair, the control program was define, by the water chemistry specifications established by Westinghouse in WCAP-7452 " Chemistry Criteria and Specifications for Westinghouse Pressurized Water Reactors", Table 3, Specification 1.3. The recommended pH range given therein, 9.0 - 10.5 in the steam generator drum, is qualified by the assertion that: Drum pH is adjusted through use of volatile amines and/or phosphate. Care should be exercised that no free caustic can exist. This can be insured by main-taining a maximum sodium-to-phosphate molar ratio of 2.6. Total phosphate concentration recommended was "up to 10 ppm" as qualified by the following: Phosphate treatment will be controlled through addition of mono , di- or tri-sodium phosphate, or mixtures of these, to maintain the sodium-to-phosphate molar ratio below 2.6 at all times. Adjustment of steam generator chemistry through additions of sodium hydroxide alone is not recommended. The water chemistry control as practiced at Robinson required operation of phosphate feed pumps when the phosphate level in the blewdown decreased below 5 ppm until that level was increased to approximately 10 ppm. The phosphate feed solution was a mixture of di- and tri-sodium phosphates, in approximately equal proportions, yielding a feed ratio for sodium-to-11
# ' 1
phosphate less than 2.6. Blowdown was not employed to regulate the water chemistry after the initiation of primary-to-secondary leakage, except insofar as continual sample flow (N1 gpm total for three steam generators) was maintained. Prior to the occurrence of leakage, generator blowdown was conducted on a periodic basis. Inspection of the 1971 and 1972 data show that pH, P0u and the Na/P0u molor ratio exhibited irregular behavior throughout the operations preceding the May,1972 shutdown with several instances of out-of-specification conditions. For example, during September and late December of 1971, the Na/P04 ratio in all three steam generators exceeded 2.6; this condition was present virtually all of 1972 precedino the repairs. The frequency of high pH, i.e. , pH>10.5, is lov. The effect of condenser in-leakage compounded the difficulty in maintaining the proper balance in the chemical environment. Table 2.3-1 contains reported analyses for chloride in the steam generators, which demonstrate that condenser in-leakage was virtually continuous during 1971 and early 1972. The chloride levels, in the range of 0-5 ppm, are not serious with respect to the limiting specification of 75 ppm; however, the chemical species accompanying chloride into the condenser (and hence into the steam generators) include calcium and magnesium, elements which act to remove phosphate from solution by precipitation of Ca3(P0g)2or Mg3(P0 4 )2, resulting in the release of Na0H. The phosphates of calcium and magnesium are readily dispersed and removed by blowdown. With the weight / weight ratios for Ca/P0g and Mg/P04 , it must be expected that 1 ppm Ca represents 1.6 ppm phosphate demand; 1 ppm Mg similarly consumes an equivalent of 2.6 ppm phosphate. Table 2.3-2 summarizes the concentration ranges for the various contaminants found in the Lake Robinson water.
In addition to the calcium and magnesium reactions tnat occur during periods of condenser in-leakage, the reaction of phosphate with metal surfaces or metal oxides represents a continuing potential for phosphate consumption. The potential for these reactions may depend upon the relative alkalinity of the solutions, but the importance of the phosphate reactions with metals or metal oxides as an influence on steam generator chemistry should not be over-looked. Combined with consumption of phosphate by reaction with metals or metal oxides in the feodwater, the effect on phosphate level is to increase the Na/P0 4 ratio and create the potential for free caustic. Compounding the difficulty in assessing the chemical environment with regard to the presence of free caustic was the presence of boron, resulting from the primary-to-secondary leakage initiated in October,1971. The effect of boric acid on the pH - P04relationship is to depress the pH measured at sample temperature, creating a misleading observation of the sample alkalinity because at high temperature, i.e. operating conditions, boric acid is virtually undis-sociated. The high temperature pH of a solution containing sodium phosphate and boric acid is expected to behave as though the boric acid is not present in effect on pH. P O 13
TABLE 2.3-1 STEAM GENERATOR CHLORIDES 1971 STEAM GENERATOR
. #1 #2 #3 January 20 .04 .16 .21 26 .07 .18 .18 i February l'/ .21 .15 .17 i March 12 .69 .64 .64
, April 26 <2 <2 <2 May 4 1.2 .49 10 20 ND ND ND June 4 .3 .2 .2 July SHUTDOWN August 26 4.07 5.17 3.01 September 3 .4 .42 .45 L 7 .67 .62 .7 17 .33 .36 .33
-October 6 .03 .06 .06 18 .05 .06 .07 27 .07 .05 .10 November 3 .08 .08 .05 l
15 .35 .2 .25 I 23 .2 .15 .04 December 3 .1 .1 .1 l 10 .15 .15 .08 14 .6 .4 .3
. ND = Not Detectable
- All Chloride values ppm -
L ! 14" -
.=. - .- . _ _ - ._ . .
TABLE 2.3-1 Continued-i il 1972 STEAM GENERATOR 1
#1 #2 #3 January 3 .3 .18 .28 14 .2 .2 .1 February 1 1.5 1.8 1.5 11 .3 .3 .4 i 21 .4 .4 .35
- March 7 .55 .50 .55
- 17 .45 .5 .5 I
30 .48 .46 .40 April 24 1.0 .8 .7-l May 10 1.5 1.2 -1.4~ j June 14 2.9 2.9 2.85 l 20 1.1 .A .4 July 12 .26. .28- .25 27 ND ND ND August 14 ND ND ND September 'l .1 .12- . 15 {
-15 .06 ND . ~. 01 2
I t l ND - Not Detectable All Chloride values ppm 15 F -. . - . .s -.,w . r - + - - , _ , ..
I l Table 2.3-2 CHEMICAL ANALYSIS OF THE SERVICE WATER FROM LAKE ROBINSON MAXIMUM MINIMUM CONCENTRATION _ CONCENTRATION Color 12 4 Turbidity 25 pH 6.3 4.6 , Free Carbon Dixoide (as C02 ) 15 4 Conductance Micrombos 24.5 13.8 Methyl Orange Alkalinity (as CACO ) 6.0 0.0 3 Phenolphthalcin Alkalinity (as CACO ) 0 3 Total Hardness (as CACO3 ) 3.5 1.7 Calcium (as Ca) 0.9 0.32 Magnesium (as Mg) 0.41 0.0 Sodium & Potassium (as Na) 4.0 1.4 Bicarbonate (as HCO3 ) 7.3 0.0 Cnloride (as Cl) 4.5 0.5 SuTfate (as SO4 ) 7.7 0.0 Silica (as S)02 ) 7.2 2.3 Iron & Aluminum 0xides (as R23 0) 3.4 0.8 Iron (as Fe) 1.1 0.07 Ammonia (as N) 0.7 0.3 Nitrates (as NO3 ) 0.4 0.0 0xygen consumed by Dichromate 27.1- 10.2 Total Dissolved Solids '48.0 22.0 l ( 16
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i 3.0 Examinations of Robinson Steam Generators 3.1 Eddy Current Examinations On May 14 the H. B. Robinson II plant was shutdown because of a reported increase in primary to secondary leakage in steam generator "A". The leakage in steam generator A increased from 55 gallons per day, which had existed from October 1971, to approximately 10 gallons per minute while returning to hot standby following a maintenance shutdown. Steam generator "A" was visually inspected following a normal plant cooldown and after draining the secondary side of the generator to within 4 inches above the tube sheet. The inspection revealed a single leaking tube on the inlet side of the generator in row 7, column 43. No other leaks were noted. An eddy current test program was initiated in steam generator "A" to characterize the defect in the subject tube and also to determine the condition of surrounding tubes. The eddy current inspection method used for non-destructive testing of steam generator tubing is discussed in and attached to this report as Appendix I. A thorough EC (eddy current) inspection of the leaking tube, row 7, column 43, revealed a through wall defect just above the secondary side of the tube sheet (approximately 3 inches). The sensitivit.y of this non-destructive testing procedure is such that indications greater than 25-30% penetration in the tube wall can be detected. It is therefore possible that indications less than this lower sensitivity level may exist in the unit. e The EC test program for steam generator "A" provided for inspection of 2029 inlet side tubes up to and through the first tube support. The area of the tube sheet tested is shown graphically in Figures 3.1-1. On site evaluation of the EC data revealed 23 tubes with indications of varying degrees of outside-diameter originated wall penetration. Table 3.1-1 lists the EC indications observed during the inspection of steam generator "A". In addition to the testing of "A" generator tubes, an inspection program was carried out in the "C" and "B" units. These inspection patterns were established to give a comprehensive inspection of those areas of the generator where indications have been observed. A total of 1588 tubes were inspected in "C" inlet with 2 EC indications observed and 864 tubes inspected in "B" with four indications noted. These indications are shown in Figures 3.1-2 and 3.1-3 and are tabulated in Table 3.1-1. The eddy current test program for the Robinson generators as included a survey of tubes in rows 3, 7 & 17 of "A" generator and rows 5 & 8 of "C" generator at a frequency of 25 Khz in addition to the standard inspection of 400 Khz. The primary purpose of the testing at the lower frequency of 25 Khz was to examine the region just above the tube sheet for indications of " crud" deposits. Evaluation of the test data is inconclusive as this technique is under development with little correlation having been established between EC signals and crud patterns.
)
TABLE 3.1-1 CPL STEAM GENERATOR "A" (Tested May, 1972)
' TUBES WITH % WALL ELEVATION INDICATIONS PENETRATIONS (AB0VETUBESHEET)
- 1. 5-41 < 50 0-3" OD
- 2. 6-20 90 i
0D 0-2"
- 3. 7-41 60 00 0-3"
,[ 4. 7-21 90 OD 0-3"
- 5. 7-27 90 I
OD 0-3"
- 6. 7-43 100 f.eaker 00 0-3" t 7. 7-45 85 i
OD 0-3"
- 8. 7-46 50 OD 4"
- 9. 7-47 90 OD 0-3"
- 10. 8-25 80 OD 0-3"
- 11. 8-29 85 OD 0-3"
- 12. 8-30 85 OD 0-3"
- 13. 8-42 75 OD 3"
- 14. 8-45 85 OD 0-3"
- 15. 8-48 75 OD 2"
?6. 8-50 75 00 0-3"
- 17. 8-51 90 0D 0-3"
- 18. 8-52 60 OD 0-3"
- 19. 9-47 80 OD 0-3"
- 20. 15-41 50 00
- 21. 16-40 75 OD 0-3"
- 22. 18-46 90 OD 0-3"
- 23. 19-48 < EO OD 0-3" 4
All the above tubes were plugged. In addition, the following tubes were also plugged. 24 7-44 2 Indication free tube removed for examination 8-l } Required plugging due to error in plugging l
. 27 8-31 outlet end of tubes 7-27, 8-25 and 8-30.
I 25
r TABLE 3.1-1 continued CPL STEAM GENERATOR "B" TUBES WITH % WALL ELEVATION INDICATIONS DENETRATIONS (AB0VE TUBE SHEET) 4
- 1. 18-41 75 OD 1/2" i
- 2. 20-44 60 OD At Top of Tube Sheet
- 3. 20-45 75 OD At Top of Tube Sheet
- 4. 20-46 85 OD At Top of Tube Sheet All four tubes plugged CPL STEAM GENERATOR "C"
- 1. 19-45 75 OD 0-2"
- 2. 15-48 75 OD 0-2" l
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R0WS f{i N0!ZLE -- l CPL STEAM GENERATOR "C" ,
+1 8
s tube eddy current probed May,1972 11230221-03 is tube clugged May,1972 O- Eddy current indications of wall per.etration . FIGURE 3.1-3 29 9
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, 3.2 Removal of Tube Specimens & Tube Plugging The tube leakage investigative program included the requirement for removal of defective tube sections. Prior to tube plugging operations, it was decided to remove sections of defective tubes A 7-45 & A 7-46 and also to remove a section of A 7-44 which had been judged to be non-defective during the on-site eddy current testing.
The procedures for tube specimen removal as well as the required tooling had been developed and tested and used for removal of tube sections from other operating generators. The tube machining, expanding and cutting operations on all three tubes were perfomed as per procedures but withdrawal of the selected portion of tube A 7-44 by hydraulic jacking was unusually difficult. Normal withdrawal is initiated with loads of approximately 3000 pounds which then normally falls to zero as movement of the tube section begins. In the case of tube B-744 the withdrawal load reached 5000 pounds and remained between 3500 and 5000 pounds for the entire length of the tube specimen. Procedurely, the tube section beginning 2 inches below the number one tube support is removed for examination. It is believed that the high temperature (150-185*F) present after secondary _ side draining dried out the tube sheet-tube interface removing any water which could have acted as a lubricant within the tube sheet crevice. When similar withdrawal difficulty occurred on tubes A 7-45 and A 7-46 it was decided to remove only the length of tubing containing the eddy current indications. This portion of_ tube A 7-45 amounted to 37 inches and for tube A 7-46 the length of tube removed was 32 inches. Since a total of 73 inches of specimen was to be removed (up to first tube support) for each tube, less then 36 inches of tube A 7-45 and 41 inches of A 7-46
- remained in the unit.
30
b Repair of the fully-removed-tube hole (A-7-44) was accomplished with an explosive plug extending through the full thickness of the tube plate. After explosive plugging, a tapered Inconel plug was inserted into the
- explosive plug and manually TIG welded to the explosive plug. In addition, a filler pass was made between the explosive plug and the tube plate surface. The remaining holes (A-7-45 and A-7-46) were repaired by inserting a 42 inch long x .710 inch diameter cylindrical Inconel 600 plug into the tube; a 7 taper on the primary end allowed the plugs to be driven into position. The plugs were then manually TIG welded to the tube and tube plate surface. It is important to note that several reductions in plug diameter were necessary to allow insertion of the plugs into the tubes. The original diameter of the plugs was .760, .015 inches less than the nominal tube inside diameter. The plugs were then machined to .740 -
.745 and again weie too large to allow insertion. Since machining to
.710 inch diameter permitted insertion, it is concluded that the nominal crevice between plug and tube is .015, maximum.
Following tube removal operations and plugging of the inlet ends of the removed tubes, twenty-three additional tubes with eddy current indications of wall penetration were explosively plugged. These explosively plugged tubes included the outlet ends of tubes A 7-44, A 7-45 & A 7-46. A secondary system pressure test after plugging revealed that 3 outlet tube ends of "A" generator had been plugged improperly. Three additional tubes in the inlet and outlet sides were plugged to correct this si tuation. A subsequent pressure test of steam generator A revealed that the explosive
- plug in tube 7-27 was leaking. Repair of tube 7-27 was accomplished by
, TIG welding the explosive plug to the tube end and then welding a solid m ,isinconel plug into the explosive tube.
s - ,
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31
" 3.3 Metallurgical Analysis Sk' etches 6f the tube sections showing locations of EC defects as determined l
at Westinghouse, and the cuts made during the metallurgical investigation are given for tube CPL-A (7-44) in Figure 3-3.1, for tube CPL-A (7-45) in Figure 3-3.2 and for tube CPL-A (7-46) in Figure 3-3.3. i f Microscopic examinations were made with light microscopy (LM) and with scan-ning electron microscopy (SEM). Some observations were made on mechanically polished transverse cross-sections of sections with and without flattening 4 in a vise to produce a 180* bend. This flattening opens up weakened OD sur-face grain boundaries near the apex of the bend. Etching was done in a 10% Br -CH 0H solution after activating with a short hcl dip. 2 3 Tube CPL-A (7-44) 1
- Two areas were examined on this tube, one in the crevice region and the other at the location of a weak eddy current indication *of a defect. Exam-ination of the crevice sample showed no ID or OD intergranular attack.
Examination of the sample containing the EC indication revealed an inter-granular attack from the OD surface of the tube to a maximum depth of 0.015" with a length of less than 1/8", Figure 3.3-4. This crack was located at a point 2 3/4" above the top of the tube sheet. t Tube CPL-A (7-45) The eddy current tes't of this tube had indicated a deep defect s5" above
- the top of the tube sheet with a length of 9/16". Examination of this area showed a defect of this length, Figure 3-3.5 and 3-3.6 Microscopic exam-inations were tade of specimens from each of three general regions; the crevice region, the defect region, and above the defect region, j
.- - A sumary of these examinations are given in Table '3-3.1. Included are
( the locations and extent of intergranular attack. Note that there is no attack in the crevice region or in the region above where there was an eddy
- Detected with laboratory equipment during initial examination of sample.
.32
t current indication. In the region of the eddy current indication, the intergranular attack from the OD surface is concentrated in an axially elongated area about 30 to 90 from the partition plate and extending length-wise between 15/16" and 1-1/2". Tube CPL-A (7-46) The in-plant eddy current defect was located by eddy current testing at Westinghouse at 2-5/8" to 4-1/8" above the top of the tube sheet. The only section examined microscopically was located in the crevice region. No ID or OD attack was detected. Comparison of Grain Size, Hardness, and Bulk Chemistry A comparison of the grain sizes of these three tubes is given in Figures 3-3.7. The grain size is quite uniform with an ASTM number of 8 to 8.5. A comparison of the grain size, hardness, and bulk chemistries of these tubes with 86 Inconel Alloy 600 heats previously studied is given in Table 3-3.2. The grain sizes and bulk chemistries of the CPL tubes are within the ranges previously observed for the 86 heats. The hardness
- readings on the CPL tubes are near the highest hardness observed on 4
the 86 heats of Inconel 600. 3.4 Microanalyses of Tube-Wall Defects Instrumental microanalytical studies were made of surface contaminants and elements within cracks or on fracture faces of sections of. tubes CPL-A (7-44) and CPL-A (7-45). These techniques included scanning electron microscope energy-dispersive X-ray analysis (SEM-XSA), electron j beam microprobe anlysis (EMA), spark-source mass spectrometry -(SSMS), i and ion-microprobe direct imaging mass analysis (DIMA) with the Battelle- ' Columbus DIMA. The SEM-XSA was used on CPL-A (7-44); the others were used on CPL-A (7-45). e 33 4 m
r-t . e o s'. P 3.
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27 I/4" 7 2 a v Figure ~3-3.1 Examined Portions of Tube CPL-A (7-44)
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t__ 7,, i I CHEMISTRY I Figure 3-3.2 Examined Portions of. Tube CPL-A (7-45) 35' _J
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- designates matching surfaces of photomicrograph. 37X t
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CPL-A (7-45), section 12D3-12D4, 5" above top of tube sheet. Intergranular cracks emanaring from OD surface extend through 95% of the vall. 100X o 39
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- .c) CPL-A (7-46), section 01A-01B.' .500X 40'
" Str.4 Aft 9 A '
-. p u
Table 3-3.1. Summary of loc.ition, depth of intergranular attack from OD surface and aieasurement method for CPL-A (7--45) . Each tabulated value is for.one crack and is the percent of the wall thickness penetrated. i t Distance (inches) Above Tube Wold and Section Number Fron Scribed 18-7/8 7/16 26-11/16 27-15/16 28-7/16 27 27-5/16 27-5/8 Line, Viewed Fren Tube 12B 12D1 '12D2 IZ13 12D4 12D5 12D6 12El ' Weld 12D 12D2 12D3 12:34 12D5 12D6 12E 12F 1 1 1 1 30 No No X 30,50 .5* No Nc 60 OD OD ,85.95 50 100, 85 OD OD 90 or or 20 30,25,30 or or 120 ID ID i.5,.5,.5 ID ID { 150 attack- attack Not attack attack 180 examined 210 Micro-240 .
.3, .3 scopically 270 .
300 , 330 360 " Measurement Method M M B M B M M M
- 13 cracks extending to a depth'of .5% of wall thickness.were observed.
M Microscopic examination (up to 200X) of polished cross-section of full rin'g. B . Half rings were squeezed and examined .st 30X.- Crack opened.on Section 12D2-12D3. X Deep creck was observed, depth not determined. 41
Table 3-3.2 Wet Chemistries, Hardness, and Grain Sizes 2 Rockwell b ASTM a C Cr Fe Ni "B" Hardness Grain Size Range for 0.02 14. 81 6.22 75.67 71 7.0 Previously Studied to . to to to to to c 86 HAPD Heats 0.04 17.0 8.25 78 12 91 10.5 1
- CPL-A (7-44) 0.032 15.1 7.96 75.2 90.8 8.5 CPL-A (7-45) 0.031 15.2 .7.93 75.1 99.5 8.0
- CPL-A (7-46)- 0.0 31 15.1 8.01 75.0 91.8 8.0 a = Includes Co content.
b~ = By conversion from the average' of3 f/ ckers readings, 10 kg load on polished and etched su* faces. lc = . Range for 8 tubes from these heats. 42
h A large short fissure opened by squeezing CPL-A (7-44), at the weak eddy current indication, contained Si, Cl, and Ca by SEM-XSA of three points on the intergranular portion of the fissure. A light-colored particule on this same face also contained P, K, and Al. All spectra also showed the alloy constituents of Fe, Ni, and Cr. The cracks in tube CPL-A (7-45) were extensively microprobed. EMA scanning of a crack located about 5" above tne top of the tube sheet showed that phosphorus extensively followed the crack; Na and Si were less extensively distributed, Cl was isolated to one point, and the signals for K, Ca, and S were approximately background. Nearer the mouth of this same crack, EMA revealed distributed Na, islands of Si, and no K above background. A part-wall crack in an adjacent section of CPL-A (7-45) was broken open to produce an intergranular fracture surface for spark-source mass spectrometry of surface contaminants. Table 3.4-1 presents the SSMS data for fracture face and OD surface contaminants on this sample, 4-11/16" above the top of the tube sheet. The data showed high levels of both Na and P, as would be expected from sodium phosphate water treatment. Major contamination levels in the SSMS data for ( species originating in the water included K, Ca, Al, Si, and C1 Other high-level contaminants from plant materials included Cu, Zn, Ti, Mn, and Mo. High levels of Li and B, which could originate from the leaking reactor coolant, were also present. Lower levels of Mg and sulfur were observed. Traces of water-originated species and system-originated species, including Ba, Sc, As, Pb, V, Co, and Nb, were detected. A third section of CPL-A (7-45) tubing about 5-S/8" above the top of the tube sheet was cut into two samples which were examined with the DIMA ion probe. The results from the OD surface revealed significant quantities of Na and K (NA > K)', together with P0 ', Cl , and OH- _ 4 l (Cl > P04 " > OH-). Other elements which were observed on the surface l 43
Table 3-4.1 CPLTubeA(7-45),4-11/16" above top of tube sheet. Spark-source mass spectrometric surface element analyses. Intergranular fracture of broken-open part-wall crack and adjacent 00. Relative; semiquantitative; units of ions /106 Fe ions Fracture Surface Adj acent Point 1 Point 2 Scan OD _ Li > 300,000 > 300,000 > 300,000 > 300,000 Na > 300.000 > 300,000 > 300,000 > 300.000 K 100,000 100.000 90,000 100.000 Mg 2,000 2,000 3,000 3,000 Ca 20,000 20.000 20,000 20.000 Ba 250 100 250 100 B 300,000 600,000 500,000 > 1,000,000 Al 100,000 100,000 50,000 300,000 Sc 400 400 400 400 Si 50,000 50.000 35,000 50,000 P 350,000 550,000 200,000 550,000 2 2,000 2,000 S+ 1,500 1,500 C1+ 13,000 13,000 11,000 15,000 Zn 100,000 100,000 100.000 200,000 Cu 70,000 70,000 100,000 55,000 As 150 100 100 100 Pb 350 200 200 ;00 Ti 30,000 30,000 30,000 30.000 V 700 700 700 700 Mn 25,000 25,000 60,000 10,000 Co -350 350 600 350 Nb 400 400 400 400 Mo 40 000 40,000 40,000 40,000 l . 44
I i included Li, B, F, Mg. The OD surface was also iron-rich in the form of an oxide layer. The element Pb was unequivocally detected at a low level (estimate < 100 ppm), and it seemed to be uniformly distributed over the surfaces. Lead was not detected in the base metal, indicating that it was a surface contaminant only. The predominant, uniformly distributed species included the alkali metal ions (Li, Na, K), raw-water species (Ca, Mg, SI), halides (F , Cl-), phosphate-related ions (P02 , P03 ), and hydroxyl (OH-) fon. Sporadic or spotty distribution of system contaminants such as Cu, Zn, Bi, and Pb were observed. Species such as Ba and Sc, which can be in natural waters, were also uniformly distributed in the analyses in which they were sought. Boron, undoubtedly from reactor coolant leakage, was pre-sent. The DIMA data confirm the SSMS data and they further show that the major SSMS species tended to be uniformly distributed in the cracks. All four microanalytical techniques offer independent evidence for the presence of numerous species which occur in natural waters and there-fore are probably the result of abnormalities, such as condenser leaks, in the nuclear steam generator feedwater chemistry maintenance. The alkali metals can produce alkali hydroxides by several hydrolysis reactions from the usually occurring anions which accompany them in untreated natural waters. Indirect evidence for one such anion, C0 3 (or HC03 -), was found by the DIMA as a C image in the cracks. The 2 DIMA also found OH ions in the cracks. Cracking of Inconel can occur in concentrated OH ions at the steam generator conditions. The analyses offer evidence which is therefore not inconsistent with a caustic stress-l corrosion cracking mechanism. The presence of high levels of phosphate which can buffer against free hydroxide production was also established ( in the microanalyses. This suggests that a caustic mechanism would require a mole ratio of alkali metal-to-phosphate which is too hign for effective phosphate buffering. These analytical results are not singly capable of either proving or disproving alkali / phosphate imbalance. D 45 i
s Also present in the analyses of material in the cracks is Pb. The Pb
- levels are trace levels compared to alkali metals, and the Pb in the cracks is sporadic and not uniformly distributed like Na+, K+, and OH'.
In summary, the analytical data indicate a complex environment of steam side corrosion products, tramp impurities, and alkaline raw water species. The only single species known to stress corrode Inconel which i was uniformly distributed in the cracks was OH . A second species also < known to stress corrode Inconel, namely Pb, was sporadically distributed i 1 in the cracks. i 3.5 Analysis of_ Sludge and Deposits i 3.5-1 Tubesheet Crud A sample of crud was taken from the top of the tubesheet in the tube lane + of CPL steam generator A. The crud sample was obtained on May 13, 1972 during dry layup of the steam generator. The chemical analysis of this j crud is presented in Table 3-5.1 i j The constituents of this crud which could be identified by X-ray diffrac- { tion along with their interplaner distances are presented in Table 3-5.2. l The multiplicity of the X-ray diffraction pattern make positive identi-- fication of both the identified and at present unidentified minor constituents very difficult. As can be seen.from both the elemental and diffraction analysis, the tube-sheet crud appears to be highly enriched in copper and zinc. Since'no copper or zinc is present in the steam generator. itself, _the source of the copper and zinc observed in the tubesheet crud could be from-corrosion of the copper-nickel and admiralty feedwater heaters tubes and condenser tubes. The presence of high ' silicon content' in the crud'could possibly.be due - to in-leakage of condenser water which at CPL is- high in silicon'(2 to 7 ppm). Any other source of such high levels of silicon is not obvious- . Y 46 '
TABT.E 3 5.1 ANALYSIS OF CRUD ON TOP OF TUBESHEET OF CPL STEAM CENERATOR A Element Weight Percent Present Ag 0.3 A1 0.43 B 0.01 Ba 0.008 Ca 0.07 4 Co 0.16 , Cr 0.65
; Cu 33.0 Fe 23.0 la 0.07 3
! K <0.03 Li 0.01 s Mg , 0.35
.Mn 0.65
, Mo 0.02 Na 0.06 Ni 7.2 P 1 ~. 5 l Pb .0.12 ,. l' . Si 3.2 Sn 0.65 i Th 0.03-Ti 0.22- , Zn 7.5 l 4 l _,_ , _ , . ~-
Table 3-5.2
- ay Qua i a ve Ann ysis of Crud ,
CoKa Radiation in CPL Steam Generator A From Tube Sheet Sample 3GA-CPL- Fe 0 Cu Zn (Cu,Zn)3(PO4 )(OH)3 "20 34 20 dA I/Io dA I/Io hk1 dA I/Io hkl dA I/Io hk1 dA' I/Io hk1 134.575 .9717 W .9692 40 555,751 1.0436 222 118.'000 1.0457 W- 5 116.995 ~1.9514 W 1.0489 40 800 __ 110.250- .1.0926 M 1.0922 60- 553,731 1.0900 17 311
'89.900 1.2782 ~M 1.278 20 220 1.282 21 110 88.000 '1.2885 W- l'.279 30 533 81.175 1.3757' W 1.342 25 103 78.550 -1.4140 W.
174.150; 1.4847. .M 1.483 85 440 172.850 1.5075 W. 67.250- 1.4164 M. 1.614 85 333,511 1.629 -8 65.000 1.6659f TJ 1.65 63 63.675- -1.6968 W. 1.687 28 102 1.705 4
~ 62.950 .1.7143 -W 1.712' 60 442 1.719 10 59.350- 1.3080 S. 1.808- -L6 200 1.826 6
-58.100 1.3434 W' 1.855 8
~57.225 1.3690 .W~ 1.872 2
'50.830 2.0868. LS 2.096 70 .400. 2.088 100 111 2.091 100 101 2.076 8 241,232
'49.575 2.134 W' 2.185 12 223
.47.775 2.'2122 W. 2.308 40 100 2.251 6 123,402 45.400 2.3195 M~ 2.335 4 410,331 42.550; 2.4668 S 2.473 53 002 2.483 25 231',032 *
~
Table 3-5.2 Continued - Sacr..c SGA-CPL Fe3g 0 Cu Zn (Cu,Zn)3(PO4 )( ")3 H 2 2 e. dA I/Io dA I/Io hkl dA dA I/Io hk1 I/Io hk1 dA I/Io hk1 41.425 2.5307 S 2.530 100 311 2.539 8 040 39.700 2.6361 W 2.640 10 202 38.675 2.7021 W 2.694 12 122,320 37.275 2.0008 W 2.771 30 230,222 36.123 2.05/0 W 2.850 6 131 35.110 2.9640 M 2.966 70 220 2.956 25 221 25 175 3.6774 M 3.642 100 121,201 24.060 4.2950 M 4.32 14 210 23.273 4.4333 M 4.48 30 120,207 21.330 4.8322 M 4.85 40 111 17.850 5.7695 M 5.92 20 011 14.300 7.1914 M 7.29 25 001 I 3.5-2 Deposited Crud Samples of deposit were taken from both a defected and non-defected tube (defects determined by eddy current inspection) from the CPL Steam Generator A. The deposit was brushed from an area extending from about 18 to about 26 inches above the tube weld with a stainless steel brush. The large area was necessary in order to obtain sufficient sample for analysis as the deposit on the tubes was slight. The crud weight averaged only about 5 to 15 milligrams per decimeter square. Some very small areas of hard deposit were located on the tube sur-faces near the bottom of the tubesheet crevice. In all other areas the crud was porous and easily remove <' by brushing. The elemental analysis of depos'ts froa both the defected (tube 7-46) and non-defected tube (tube 7-44) i: prasented in Table 3-5.3 and the X-ray diffraction analysis of the deposit constituents is presented in Table 3-5.4. Only the sample of deposit brushed from the non-defective tube (tube 7-44) is included in the table as the amount of material brushed from the defected tube (tube 7-46) was insufficient for any accurate analysis. J The deposit on the tubes just above the tubesheet appeared to be very rich in copper which is confirmed by both the elemental analysis and X-ray diffraction analysis. The high nickel and chromium observed by analysis in the deposit is probably due to contamination of the deposit with the base metal during sampling. An examination of the deposit analyses (Tables 3-5.3 and 3-5.4) show the following: 1 There is a rather high concentration of aluminum in the deposit , sample which may arise from system materials or from the aluminum found by analysis in Lake Robinson water.
- 2. The presence of boron is attributable to boric acid which leaked through the steam generator tubes from the primary to secondary system.
~
l
. . _ ~ _ - _ _- . . . . -
l-
- 3. - The iron content of the deposit is low due to the high concentra-
! tions of nickel and chromium contaminants.
- 4. Copper and zinc appear as significant concentrations in the deposit and may arise from system material corrosion.
- 5. The sodium and potassium content of the deposit is significantly high but may arise from the in-leakage of Lake Robinson water which contains sodium, potassium and silicon at about the ratios observed in the deposit.
- 6. Phosphate is present in.the deposit at about the level to be expected in a steel system operating with phosphate chemistry.
- 7. X-ray diffraction patterns from the deposit sample show the presence of Fe340 , Cu, FeP0 4 Ca (P0 )30H (hydroxy apatite), NaCaB0 and 5 4 3 Al230 . 4SiO 2"20 with the possibility of (nap 0 )6 3 as another con-stituent. The presence of the silicate and apatite would indicate that some source of raw water was fed to the steam generators. The most likely source would be a leaky condenser which would permit entrance of Lake Robinson water into-the steam system.
) Measurements of pH were made on the tube surface deposit. The pH values-determined on tube surfaces from within the tubesheet were about 12 to 12.5. The pH values determined on the tube surfaces just above the > ! tubesheet (up to 4 inches above) averaged ~ about 12.5. The pH determina - tions made at greater than 4 inches above-the-tubesheet were all about . 10 to 12. The high pH values observed on the tube surface in the crevice area
- and at the top of the tubesheet 'would indicate that eitherlthe' crevice area.had not been washed or that the deposit in the crevice area was ~
only slightly permeable by water.
- _ ._2 ._
i L I (.
TABLE 3-5.3 ELEMENTAL ANALYSIS OF DEPOSIT BRUSilED FROM CPL STEAM GENERATOR A TURE 7-44 AT 18 TO 26 INCHES ABOVE TUBESIIEET Element Weight Percent Present Ag 0.006 Al 2.0 B 1.3 Ba 0.2 Ca 0.1 Cd 0.03 Co 0.11 Cr 11.1 Cu 4.7 Fe 12.0 In 0.03 Ki 0.95 Li 0.08 J Mg 0.5 Mn 0.2 Mo 0.61 , Na 3.0 [ Nb 0.12
- f. Ni 34.5 *
! - P 0.8 1 Pb 0.27 { Si 5.5 Sn 0.6 Th 1.8 1
.t Ti 0.5 s ; Zn 1.0 ]
Zr 0.02 I t t l k
- i i*
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m_. 6
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e
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l @. 411*A'1MJ,li"
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# I#I e h3L d I#I e hAL d I#I # I#I 4 I#I e h"L d I#I e hEL e d I'I e
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h"I' e e 20 l.271 75 1.24 10-20 90.525 1.262 W 1.204 le 622 20 1.292 75 1.20 le 89,.40 1.277 W l.279 30 $33 1 . 2 73 502 1.49 70 1.474 25 1.49 20-40
} 14.1750 1.48 4 m 1.433 e5 40 1.474 11 1.$0 30 1.500 2S 20-40 F2.9125 1.904 y 1.503 9 214.421 1.527 73 1.52 10 l 72.2750 1.918 W '.330 5 331 Sc1,206 1.s7 to 1.57 s 1.98 w 1. 587 3 l 69.6250 1.4 s 40 1.62 30 1.62 10 20 67.34?5 1,613 v 1.614 35 333.511 1.611 7 113 1,64 25 1.46 30 1.65 50 1,64 20-40 65.9250 1.64% v 1.6% 9 322,223 g
v 1,712 to 422 1.722 20 004,411 1.72 75 h 42.9500 1. 7 % 4 40. *D 5 1.765 W 1.7% 15 402.303 410 1,77 2$ 1.78 40 1.770 7S
, 60. % 23 l.790 v 1.750 11
' 59 4$00 1.79. M 1.80 50 1.836 2C 321 1.834 46 1.602 ' 75 59.M47 1.807 W 40 213 1.84 10 1.057 75 1.83 40
- 57. 625 1.853 n 1.e41 Sto 1.88 5
!6 e425 1.880 v 1.871 5 15 312 1.91 10 56.3062 1.B97 W 1.090 1.%3 30 222 1.914 75
. 55.0062 3.938 R y 1.97 10
, 55.4125 1.992 8 53.1000 2.002 9 2.000 S 203 2.040 400 2.01 30 2.01 23 i '2.2000 2.016 b 1 2.07 600 2.065 7 113 2.088 100 111 2.07 21 10 2.090 h 2.096 70 f $0.6931 p 2.134 3 302 2.14 10 2.14 10 20 49.4175 2.141
.4.6667 2.1E U 2.148 9 311 2.19 50 2.17 25 6a,3437 2.185 n v 2.228 1 221 2.25 Il 47.00J0 2.246
[ 46 4637 2.260 n 2.262 to 310 u 2.296 7 212 2.M 75
- 43. woo 2.324 2.40 to 2.40 to 43 9375 2.3 91 v 2.419 10 222 2.44 30 43.3062 2.426 M 42.4500 2.472 W 2.52 10 2.52 to 41.5125 2.526 x 2.530 100 111 2.549 W 2.525 5 301 2.59 to 41.1125 2.M 100 43." WS 3.111 M 4.4%2 2. ; *2 n 2.631 25 2G2 2.64 10 19 ' 37 ?.632 W h
2.67 10 19.*.412 2 6eJ 8 l l 30 % 3? 2.M4 W
?9. lii! 2 .'26 W 2.720 to 300 2.74 71 4' e437 2.76J W W 2.775 60 112 2.79 10 3 bt ? > 2. 7 h 5
- s. 8754 2 811 4 2.814 100 211 2.88 100 94 .1 6 ?! 2.8f4 W 2.92 75 35.6$25 2.9% sg 2.95 100 15.?'i00 2.952 W 35 dig? 2. 976 M 2.M4 70 220 3.06 100 14.'437 2.998 M 3.1? 11 102 3.18 73
- 32. 137 3.179 w 3.32 80 3.34 20-30 11.3675 3.312 W 33.7e2$ 3.375 W 3.44 40 002 3.45 OS 30.;500 3.441 S 2
g 3.51 1 201 29.4750 3.513 2n uS75 3.5% M 79.493? 3.g)e g 27.t882 3.714 n 9 111 3.87 5 26.7167 3.874 u 3.88 9 200 4.15 20
- 21. ele 7 6.132 p 4.87 4.57 40 21.4n00 4.%77 S, 4.61 90 D 3542 4.616 W 21 9437 4.611 m 40 111 4.72 3 110
'8,5112 5.559 w 5.26 1 sol 14 g,09 7.M4 W 11,7s79 8 .7 21 s
-F}e e
-o oea. aev *. e4 e-m ~ .w .e e. s m om w- --
- 4. PROBABLE CAUSE OF TUBE FAILURE Eddy current examination in each of the three Robinson steam generators showed 28 indications in addition to the leaking tube in "A" generator.
These indications were confirmed to be tube wall defects after laboratory examination of the three tube samples removed from "A" generator. These defects, described in Section 3, are characterized as intergranular cracks originating from the tube 0.D. surface. The defect indications were all found to be located at, or slightly above, the upper face of the tube sheet. Chemical examination of sludge deposits on the tube surface and of material found in the recesses of the crack indicated the presence of caustic, as sodium and potassium hydroxides, plus other ordinary con-taminants and additives. The inventory of elements and compounds found during analysis is described fully in Section 3. The most detrimental species of those mentioned above is the alkali hydroxide, which in concentrated solutions is known to attack Inconel as it does stainless steel and other alloys. It is possible that the caustic attack was enhanced by the presence of another element, such as lead dissolved in the concentrated hydroxide solution, but it is believed that caustic was the primary aggressive agent and was sufficient to have caused the damage. The process that develops concentrated solutions of alkali hydroxide is described by H. M. Rivers
- where it is illustrated that with a given source of free alkali hydroxide present in _the steam generator water, the concentration can increase to the limit defined by the amount of available superheat at the specific location of interest. In the hot leg crevice region of the steam generator, the available superheat, given by the-
*H. M. Rivers , " Concentrating Films: Their Role in Boiler Scale and Corrosion Problems", Engr. Soc. ot~ West. Pa. Water Conf. , Proc. , 131-145(1951).
difference in reactor coolant hot leg and steam side saturation tempera-ture, about 70 F, can lead to significant hydroxide concentrations as shown in Figure 4-1.1. The same concentrating process will occur in any location where boiling occurs and a continued supply of boiler water makes up for steaming, with an attendant accumulation of dissolved chemicals. While the crevice region in a steam generator is one location whare a maximum concentration of caustic could occur, there are many locations in the steam generator where solute accumulation is influenced by the rate of steaming, the hydraulic flow rate across these locations, and physical dimensions determined by structural design and accumulated deposits of contaminant sludges. In the case of the Robinson tube defects, which were all located at just above the tube sheet, accumulation of caustic occurred above the crevice region, that was brought about by the following contributing factors:
- 1. A source of caustic. Water chemistry analyses during plant operations showed the presence of free caustic during much of the time, as defined by a sodium to phosphate molar ratio above 2.6. Figure 4-1.2 provides the Marcy-Halstead basis of determining molar ratios. As the boiler water containing free caustic was fed to a zone where concentration occurred, sodium phosphate would deposit out as a solid, leaving the residual excess sodium hydroxide to concentrate witnout a corresponding amount of phosphate buffer.
- 2. A physical geometry permitting solute accumulation. As sug-gested by the data obtained with the 25 kHz Eddy Current test, there appears to be a concentration of sludge located on the tube sheet in the immediate area of the tubes, more or less in the center of the hot-leg tube bundle. Such a layer of semipermeable material could provide a physical arena where only limited washing of tube surfaces by the boiler water would occur.
l l
- 3. A low hydraulic flow rate. In April,1971, the Robinson steam genera-tors were modified in order to reduce the amount of exit steam moisture entrainment. The modification consisted of lowering the generator orifice plates, thereby reducing the amount of internal circulation of the boiler water. In so doing, the inlet water velocity across the tube plate was reduced from 4.3 ft/sec to 2.7 ft/sec. This reduction contributed to the accumulation of sludge on the tube sheet which harbored the concentrated chemical solutes resulting from steaming operations.
- 4. Factors that aggravated the conditions. Several items are known to have contributed to the poor chemical conditions in the steam gen-erators, which are:
- a. Only a very low blowdown rate was practiced, particularly during the first months of 1972. This permitted accumula-tion of corrosion product oxides and other impurities within the steam generator.
- b. The existence of condenser leakage, coupled with low blow-down, significantly affected the control of phosphate chemistry, leading to the accumulation of free caustic in the boiler.
- c. Phosphate additions were periodic, rather than continuous, which led to extreme variations in chemistry control. Under these circumstances, phosphate was consumed by reaction with corrosion products and condenser cooling water leakage without continually maintaining a margin against the formu-lation of free caustic.
Thus, the cause of steam generator tube failures in the H. B. Robinson II plant is attributed to intergranular stress assisted corrosion, where the corrosive agent was determined to be alkali hydroxide (caustic) that was concentrated to damaging levels on the hot leg side of the tube bundle, Concentration of the caustic which existed in the steam generator water was made possible by the ex-istence of a sludge layer that accumulated on the tubes or tube sheet as a re-sult of low circulation flow and low blowdown.
Curve 586789-A 70 i i j i i l l l l l l 1 i 60 --- - 100 psia - C 3 50 770 psia . a 3 .' (n .
.5 -
g40 - (1000 psia ~ z
's 30 - -
- c. / // t 2m 5 20 -
10 - - 0- I I I ! I I ! I I I i I - 0 10 20 30 40 50 60 70 80 90 100 110 120 130 Degrees of Superheat, F Figure 4-i.I -Concentration of pure NaOH solutions in superhea'ted crevices at selected total pressures i l i
i 693-13 l4 jf-J . f pL 4 t'
....{ p f
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FRd4 MARCY AND HALSTEAD i . i 13 ' I i i I T I
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7.5 1 10 ~100 1000 10,000 PHOSPHATEINSOLUTION(ppmP0) 4 pH of Orthophosphate Solutions Having Various Mol Ratios of Sodium to Phosphate Figure 4-i.2 - O
i 5.1 Water Chemistry Control . . . . . . Consequent to this investigation, revised specifications, operating guidelines and chemical injection systems were instituted fcr the
~
steam generators chemistry control. Speci fications:
- 1. Phosphate level be maintained s25 ppm.
- 2. Operating Guides: ' Continuous feed of disodium phosphate balanced by continuous blowdown. The feed ratio may be varied in order to
, .. achieve a Na/P0 4 ratio in the blowdown
~
close to feed solution ratio. r '
, 3. Chemical Injection System:
- a. Interconnection between existing phosphate injection line end auxiliary feedwater line at a point downstream of the auxiliary feedwater punp. A similar connection for hydrazine,
- b. Make-up water connections to the chemical feed tanks.
- c. Separate chemical feed systems for each steam generator.
Crossovers may be provided for temporary use in the event
~
of pung failure. After the tubes were plugged, the plant was returned to service. During iniUal startup after the repairs, Westinghouse Chemistry personnel assisted in establishing proper, stable chemistry control. As expected, phosphate control was established after a short period of high phosphate cons umption. However, the presence of condenser in-leakage, resulting from a steam dung to the condenser, obscured the behavior of the N4/P0 4 ratio from analytical observation; calculation of the Marcy-Halstead ratio from the pH - P04relationship showed that, even in the presence of extensive condenser in-leakage, effective chemical control could be maintained. The margin, against the formation of free caustic, provided by the 20 - 30 ppm P0 4 level meintained during the startup period was f I O 1*
completely adequate to protect the heat transfer surfaces. Subsequent operations by plant personnel have been relatively unconplicated, resulting in a markedly improved chemistry control. It can be seen by examination of Figures 2.3-4, 2.3-5, and 2.3-6 that while pH and PO 4 levels have not shown exceptional stability in terns of concentration changes, the Na/P0 4 molar ratio exhibits relatively little variation. l t
~
5.2 Improvement in Steam Generator Circulation The original design of the Robinson units provided for sufficient penetration by the feedwater entering the tube bundle to avoid stag-nation and surpress boiling near the tube sheet. Modifications which were made to the downcomer flow resistance plates to improve the moisture separator performance by reducing the circulation ratio
- resulted in a reduction in the tube bundle inlet crossflow velocity. A region of low flow creates no significant detrimental effect in itself, however an accompanying increase in average tube bundle steam quality due to a reduction in circulation ratio increases the tendency for boiling near the tube sheet. As a result, any contaminants introduced into the steam generator may tend to concentrate in this region.
In order to eliminate these effects, the circulation ratio of the Robinson units was restored to design conditions concurrently with the tube inspection and repair. This was accomplished by raising the modified downcomer resistance plates 3 inches. To alleviate the detrimental effects of increased circulation on moisture carryover, the swirl vane outlet orifice was reduced frem 43.75 inches to 36.0 inches by the installation of new orifice plates.
- Reduced from a calculated value of 3.15 to a calculated value of 2.0.
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i APPENDIX I Eddy Current Inspection - Non Destructive Testing of Steam Generator Tubing The Zetec MIZ-4 Eddy Current System is used to inspect steam generator tubing. The MIZ-4 system uses eddy currents as the probing media to measure variations in effective conductivity and/or penneability of the tube being tested. An alternating voltage is impressed across the test coil. The magnetic field developed by current flow in the test coil causes eddy currents to flow in the tube wall. The corresponding magnetic field caused by eddy current flow in the tube wall is out of phase with the field developed by the current in the test coil. Since these fields tend to cancel one another, the coil voltage is decreased in proportion to the magnitude of eddy current flow in the test piece. The magnitude of eddy currents in the test piece, thus the coil voltage, is dependent on the electrical properties of the tube being tested. The electrical properties which affect the flow of eddy currents are permeability and conductivity. In non-magnetic materials, such as inconel and 300 series stainless steel, conductivity decreases due to a discontinuity in the tube wall, the coil voltage increases in direct relation-ship with the effective conductivity change. Thus, the amount of increase in coil voltage is related to the size of the discontinuity. The coil voltage is sinusoidal, thus it can be dascribed with a single vector having magnitude and phase. The MIZ-4 system provides a method for reading out two quadrature components of the test coil vector. A differential test coil probe operating at a frequency of 400 Khz is used for the inspection. The application of these parameters allows a rapid inspection with optimum wall penetration and sensitivity for the steam generator tubing. A 30% loss of wall x 10 square millimeters can be reliably detected in noise free tubing.
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- 11. B. Robinson No. 2 Carolina Power and Light Co.
Report No. DC-67 May 23, 1970 1 Prepared for: U.S. Atomic Enorgy Commission
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Page 2 of 24 l Distributiont Copies: AEC Regulatory _ Organization Division of Compliance Technical Programs - Washington (2) ) L. Kornblith, Jr., H. R. Denton Region-II, Atlanta (8) W. C. Seidle, U. Potapovs AEC Contracts Division (~) Chicago Operation Office (1) H. N. Miller, E. Halman r Parameter, Inc. (1) i
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Paga 3 of 24 f'] (J R Table of_ Contents: Pace: 1 Introduction and Scope of Inspection 4 II. Summary of Findings 7 III. Discussion of Inspection Items 9 IV. References 23 V. Appendix 24 Exhibit-A Stresses for Safety Valve Nozzle As Presented by Ebasco, at 5/15/70 Meeting Exhibit-B Schematic Arrangement of Main Steam Line Safety Valves as Presented by Westinghouse at 5/15/70 Meeting Attachment No. 1: Sketch No. 1, General Arrangement = Safety Valve Nozzle, 26" Main Steam Line Attachment No. 2: Sketch No. 2, Safety Valve Nozzle Fabrication Detail Attachment No. 3: Sketch No. 3, General Contour of Failure, Safety Valve Nozzle Attachment No. 4: Stress Analysis of Failed Main Steam Safety Valve Nozzle -H. B. Robinson No. 2 Carolina Power and Light Co. i kd.[.)D(@ti_'M_TE?ir"Jfl Eb"# "" "-' " ~
!s/' Paga 4 of 24 6t I. Introduction and Scope of Inspection 9) At the request of the Division of Compliance (Mr. W. C. Seidle, Senior Reactor Inspector Region II and Mr. D. L. Pomeroy, Technical Support Branch, CO-Hg), the writer attended a meeting held at Ebasco Services, Inc. offices in New York on May 15 to review the recent failure of a safety valve in the main steam line of the H. B. Robinson No. 2 plant at Hartsville, S.C. The writer accompanied Mr. W. C. Seidle, Senior Reactor Inspector and Mr. U. Potapovs, Metallurgical Engineer, both of the Region II offices at Atlanta representing the AEC in this review. The Robinson-2 plant is a 663 MW(e) Pressurized Water Reactor (PWR) generating station being built for Carolina Power and Light Co. under a turnkey contract by Westinghouse Electric Co. Ebasco Services, Inc. is Designer and Constructor. The failure of the safety valve nozzle has been assumed to be metallurgical and/or stress related and investi-
/7) gations have been pursued in these areas by Westing-house and Ebasco respectively. A previous meeting (not attended by the writer) had been held at Westing-house Pittsburgh and was concerned with preliminary metallurgical findings and planned investigative effort. The prime purpose of this meeting was to review further information on the metallurgical results and to obtain the results of Ebasco's design-analysis.
Mr. Potapovs is expected to report all metallurgical aspects of the failure. In this report, the author will document the meeting presentation related to design and summarize Parameter's evaluation of the safety valve connection in terms of stress levels presumed to occur at time of failure. In this latter respect, the basis will be an independent estimate of stresses prepsred by our Mr. W. J. Foley which is appended as Attachment 4 9
i Paga 5 of 24 I. Introduction and Scope of Inspection - Continued Note: Preparation of an independent evaluation of stresses in general parametric terms was agreed upon with Mr. Seidle after the analysis summary presented by Ebasco during the meeting was not complete or conclusive. The Parameter analysis would include the effects of pressure and thermal discontinuitics, and the possible superposed moment due to the safety valve discharging. Other agenda items of the meeting include a general review of the incident, a discussion of the status of the review of the remainder of plant piping, the status of a " design fix", and the status of the investigation of the plant transient (primary system) due to the incident. These items are not considered within the scope of this report except to make some comments as to the criteria for a satisfactory fix. The following personnel participated in the meeting held on May 15, 1970 at Ebasco Services, Inc., ( 2 Rector Street, N.Y. N.Y.: (Titles or function may be approximate.) Carolina Power and Light Co. : H. Banks, Resident Engineer Westinghouse Electric Co.: T. Erion, Project Manager, Robinson-2 R. Cunningham, Research - Pittsburgh Ebasco Services, Inc.: { 1 J. Scarolla, Projects Manager (part time) R. Pulito, Project Manager, Robinson-2 l i A. Rossi, Staff 'O t/ i 15 TM E ~ Yp " e- I '
3 Pag 2 6 cf 24 1 Introduction and Scope of Inspection - Continued I. Ebasco Services, Inc.: continued l B. Mazo, Metallurgist i L. Labriola, Supervisor - Stress Analysis i S. Boone, Stress Analyst 1 U.S. Atomic Enercy Commission, Division of Compliance: W. C. Seidle, Senior Reactor Inspector - Region II, Atlanta U. Potapovs, Metallurgical Engineer - Region II, Atlanta R. A. Lofy, Consulting Engineer - Parameter, Inc., Elm Grove, Wis. (~h (/ The writer's comments on the description and various details of the failure incident are included with the narrative report of the meeting in Section-III, Discussion of Inspection Items. Significant conclu-sions are reported in the Summary of Findings, Section-II following. (Note: All notes made, computer printouts and calculations performed in connection with this assignment are maintained on file by PARAMETER and available for future reference as required.) G
Page 7 of 24 II Summary of Findings:
- 1. The safety valve nozzle was not analyzed for the load imposed by the jet reaction of the steam during
- discharge of the valve. The stresses due to this reaction can be expected to comprise a large fraction of the total maximum stress occuring when tho valve
" pope". If the valve popped, the resulting moment-reaction coupled with pressure end load and struct-ural discontinuity stresses, can be considered a major contributor to the failure.
If the valve did not pop, it is not possible to explain the overload failure in terms of pressure, weight, wind load and structural pressure discont-
. inuity stresses. Based on estimated temperature gradients, thermal discontinuity stresses across the failed area turned out to be nominal.
Please refer to Attachment #4 for a summary of stresses calculated in the trial analysis performed by Parameter, Inc., and conclusions regarding the probable stress related basis for the failure.
- 2. It has been determined that an axial tenelle stress on the order of 60,000 psi can exist in the~ nozzle wall under the steam discharging condition.
(Calculation of this stress required making ~ some assumptions regarding the reaction , forces as described in Attachment #4.) This approximate stress level, without a stress' ' concentration factor is sufficiently high to be a prime contributory cause of failure without major metallurgical defects being present. The Mill Test properties of the material have been quoted at 48,500 psi Yield and 75,700 psi Ultimate. Specification minimums for this material are-35,000 psi and 60,000 psi respectively. The margin in material strength available and the uncertain-ties of assumptions made in the trial analysis do not preclude the possibility-that~a minor ~metallur-gical defect could have been a contributory factor.;'. ; assuming the steam discharge mode. O i. V J 4m _A~pepgg AnnaM_(2AB --
Page 8 cf 24 f I. Summary of Findings: continued 3 Although the Power Piping Code (B31.1.0-1967 assumed to apply) does not appear to treat safety valve dis-charge forces directly, it is the opinion of the writers that omission of this force in establishing the design is a violation of the general criteria of this code. (Para. 101.5 covers " Dynamic Effects" and para 101.5.1 " Impact" specifically in stating that "Inpact forces caused by all external and int-crnal conditions shall be considered in the piping design.") The valve discharge would most certainly be consider-ed a dynamic effect and can be classified as an impact force. G 0
Pagn 9 of 24 III. Discussion of Inspection Items In introducing the meeting attendees to the back-ground and circumstances under which the steam line l safety valve failed, Mr. T. Erion of Westinghouse i presented a brief description of the plant arrange-ment and tests in progress at the t'me of failure. Each of the three main steam lines A, B and C (or 1, l 2 and 3) from the Nuclear Steam Supply System (NSSS) contain one power operated blow-off valvu and four (4) spring loaded safety valves set to relieve at 1075, 1105, 1125 and 1140 psig respectively. These valves are located outside of the containment and upstream from the main steam isolation valves,which were closed, and check valves. Exhibit-B shows the relative position of these valves schematically as presented to the group on the blackboard. The failed nozzle was associated with the highest setting safety valve (1140 psig) located in line C(1) as indicated on Exhibit-B. The safety valve was identified as a Crosby Style No. HCU-65W (Ref. 1). Two sizes are , O used in the system , 6" (inlet) by 10" (outlet) and 6" x 8" . The failed valve was believed to be a 6" x 8" as stated by Mr. Erion. l (Note: Inquiry Memoranda (References 1 and 2) were made available to the writer prior to the meeting and contain pertinent data regarding test conditions (pressures and temperatures) . before and after the incident, and records of prior inspections.) The Robinson plant was in the last stages of hot functional tests at the time of the incident. Mr. Erion explained that the valve testing was-left as one of the last items in their test program. Crosby personnel had been called in for the tests to make any adjustments to valve settings that might be required. They had completed tests on steam lines A(1) and B(2) and were working on C(3) at the time-of the failure. O _ m _ _ _ _ _ _ _ _ _ _ _
Paga 10 cf 24 l l III. Discussion of Inspection Items: continued i Oq i l The test procedure was described approximately as follows: With steam pressure below the valve dis-charge setting (say 900 psig) a pneumatically operated lift device is attached to the valve and enough pressure applied to simulate the differential between the actual steam pressure and the blow-off pressure. The pressure to the lift device was applied by means of a nitrogen bottle and regulator. The object is to add just enough pressure to get the valve to " simmer" or leak steam. This is the calibration point. However, it is ch'racteristic a of the valve to " pop" wide open when only slightly higher than
" simmer" pressure is reached. As an example of the delicate balance, between " simmering" and " popping",
of the eight (8) valves previously tested by experi-enced Crosby personnel in this series, three (3) simmered only, but five (5) popped. The writer asked how long the system was at elevated (71 temperature and pressure prior to the failure. Mr. Erion replied that the hot functional tests had lll been in progress with the cteam lines heated for nine (9) days. This steady state condition precludes the need to consider temperature and pressure transients in our evaluation of the failure. However, steady state conditions do not imply that uniform temperatures exist in the area of the valve to header connection because of the termination of the insulation and change in sections and surface areas at this point. Mr. Erion went on to explain how, immediately after the incident, an investigation team was set up to conduct the:
- 1) Design review, !
- 2) Metallurgical investigation, and 3) To evaluate effects of the transient on the primary system.
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} III. Discussion of Inspection Items: continued (While this report is primarily concerned with the design review aspects of the failure, discussion of the transient will be reported and observations made to the degree that the writer participatec in this discussion.)
In discussion of the progress of the test of the valve in question, it was established that the pneumatic , lift device was being used and they were starwing to go up in pressure. It was the reported opinion of test crew members that the valve did not " pop". However, there seems to be conflicting opinion in this respect. Observers report hearing two distinct sounds - one could be the valve popping, the second ) could be the escape of steam upon rupture of the j nozzle. On the other hand, the first sound could l ! have been associated with the break in the nozzle followed by the discharge report of steam. It was stated by Mr. Cunningham of Westinghouse that exami-
~
nation of the valve showed grease in the area of the 1 I valve stem and paint in the exhaust section which did not appear to have been scrubbed away by exhausting steam. (This valve was reported not to have simmered during previous tests.) 1hus the feeling that the valve did not pop. Mr. Cunningham did point out l however, that the grease was not in the direct path of the steam and the general appearance was not a l great deal unlike that of other valves which haf popped. It is the writer's opinion that, at best, this evidence as presented is inconclusive. If the valvo popped and failed soon afterwards, it would not have experienced a great deal of steam flow. If conclusions are made on the basis of these indicatiens in the future, they should be developed from a more thorough study of the history of operation of this valve and comparison with other valves of known usage, and confirmed possibly with tests under controlled conditions. Whether or not the valvo popped is cf course an 5 important consideration in analysis of the cause : I of failure, particularly if predominantly stress W related. Conclusions as to this probability can be drawn from the trial analysis presented in Attachment No. 4 s ca w w,,qat:m e n--~_____ e e 19 - - - ~~ ~
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PagJ 12 cf 24 ( III. Discussion of Inspection Items: E ) continued The writer asked whether an estimate of the tempera-ture of the valve was available. Ebasco personnel indicated that one could put his hand on the valvo in the upper area, but they had no data on the overall temperature. profile or the specific temperatures in the transition area adjacent to the no le. It was
, pointed out that actual in-service surface tempera-tures of these valves would be relatively easy to obtain by "Tempil" stick or tPermocouple.
Valve body temperature could be an important consideration in analysis of the failure area. For our trial analysis of Attachment #4, we have estinated a nunber of temperature profiles to establish a trend of thermal stress versus temperature gradient and to determine its significance. The main steam header was described as being lagged r with canned Unibestos type insulation. Study of photos of the installation showed the insulation as terminating near the top of the taper at the lower
.cnd of the valve body. The fabrication design of the
/ valve to header joint was described by a layout dis-played during the meeting. On the basis of notes and sketches made during the meeting, this design has been drawn approximately as displayed in our Sketch-2, Attachment-2 The weight of the valve was given as 950 lb. without the stack and 1300 lb. with the stack. Its general arrangement with respect to the stenn line is shown on Sketch-1, Attachment-1 There is no external support to either the valve body or the elbow-stack assembly. i The header and valve assembly was hydro-tested at i 1356 psi during a 4 hour pressurization cycle l' according to Mr. R. Pulito of Ebasco. This particular valve did not weep during hydro. (Valve presumably blocked shut at test pressure.)
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Pagn 13 of 24
\ III. Discussion of Inspection Items: continued 2
r The header and transition piece (nozzle) were designed : l by Ebasco Services andjfabricated by B. F. Shaw Co. ! l The nozzle was made from A106 Gr. B Schedule 80 seamless pipe. The weld between the valve body and the nozzle in the reduced section (Schedule #40) was made with the aid of a backing ring Icft in place. This weld was made by field personnel under Ebr,sco's ' 4 construction contract. " The field weld was not stress j , relieved. Note: While Ref. 1 reports stress relieving in j i accordance with code requirements, this section, g being under 3/4" wall thickness, would not , ) require stress relief. i It is difficult to predict the magnitude of residual stresses due to welding in a case like this. An observation can be made however that the hoop restraint or pinching effect of the residual stresses in the weld tends to add to ("' T the pressure and thermal discontinuity stresses: already present at severe changes in section in the failure area. The' backing ring, which is assumed fused to the weld provides even further restraint under pressure and temperature. Mr. Cunningham of Westinghouse _went on to report the results of their metallurgical investigation to date. -' He explained that be had been called in fron the ! Westinghouse Research organization to. work with the project group'as r. more or less independent consultant. ' The character of the failure was described-by-Mr. Cunningham at the board essentially as reconstruc-ted on.our Sketch-3,1 Attachment-3 From a design--
. analysis point of view, it suffices to say that hei f g' described the failure as a typically tensile break. }:
and_ reported that no abnormalities were yet found'in ' the physical properties of the material. -[ F N 4 g-i / A
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Paga 14 of 24 "T III. Discussion of Inspection Items: continued They believe that a crack started from the inside (not confirmed) at initiation of failure from the west edge which progressed symetrically about the E-W center line to the cast edge. (Sco Attachment-3). The inner edge of the break followed the transition line between the Schedule #40 diameter and the tapered section, and thus there was no opportunity to establish the tool radius of this corner. However, Westinghouse plans to measure a tool radius for the adjacent nachined surfaces which might be an indica-tor to that which existed at the transition line. Some necking-down of the failed section was observed. Westinghouse reported that single wall radiographs showed no rejectable defects. Meta 11ographic examina-tion showed only some normal longitudinal stringers in the pipe. The preliminary conclusion by Westinghouse is that there are no materials problems and that an overload (~N type of failure is indicated. The final metallurgi-cal report was expected to be issued within a few weeks of the meeting date. Questions and answers during the discussion of the failure revealed a consensus that the nozzles cr transition pieces were machined as detail parts in a lathe (vs. boring after assembly to the header) and that the Schedule #40 counter-bore and transition taper were made in the came setup. If this is the case, as would be expected, we should not find any offset or eccentricity between the counterbore and taper which could act as an additional stress concentration factor and discentinuity. It was asked whether other nozzles could be inspected to verify this fact. Mr. Banks (C. P. & L.) said it was not possible to view this area adequately with the valves welded in place. However, it does seem possible to the writer that some in-situ inspection could be performed by removing the valve mechanisms and viewing with a boroscope or similar device. If it is decided to leave the other valves in place, such an inspection to reveal any incipient flaw might be well warranted. C4 m i m .. - . m es ~ m ~ M ~ ~" % " = '= = '
Pcg2 15 cf 24 III. Discussion of Inspection Items: continued During the afternoon session, Ebasco stress analysis personnel presented the results of their design-analysis as recorded on Exhibit-A attached. The stresses as displayed initially on the board by Mr. S. Boone are shown in the original columns. Revisions to these numbers,as evolved from modification of the assump-tions upon which the analysis was based by Mr. Pulito and FT. Labriola are offset and shown in parentheses. Thc meaning and applicability of both the original and revised numbers can best be explained by summari-zing the results of somewhat probing inquires by the writer and responses obcained as follows. As the mode of failure appeared to be related to longitudinal loadings, Ebasco tabulated only longitu-dinal stresses at 900 psig, 1140 psig and 1356 psig. The 900 psig valve was the approximate pressure at the time of failure, 1140 psig was relief setting of the valve and 1356 psig the previous hydro-test temperature. The stresses at 900 psig and 532 F (?) operating pressure are of greatest interest to us in this evaluation. The primary longitudinal stresses were due to internal pressure, weight of the valve, and a wind loading of 30 mph. estimated at the time of failure. The primary axial stress is tensile and its constituents self explanatory. Two secondary stresses (bending through the wall thickness) were calculated:
- 1) " Bending Stress Due to Eccentricity" (offset of mean diameters of walls at change in section) and 2) " Bending Stress Due to Step in Thickness" (this ic a pressure discontinuity stress at the change in section of the wall).
The mathematical model for the " original" stresses is shown on Exhibit-A and depicts an abrupt change from the Schedulo #40 (.28") wall section of the 6" diameter nozzle to a wall section of approximately 1 " in the valve. This treatment with a 90 step at the change I f A ** Y
Paga 16 of 24 III. Discussion of Inspection Items: continued in section is harsh and the analysis should be conser-vative. Bending stresses are added arithmetica11y and summed with axial tensile stresses to arrive at a conservative maximum stress of 24,070 psi (the original case). As the opposing effect (opposite sign) of the bending stresses due to eccentricity and change in section are neg1ceted and the maximum eccen-tricity at an abrupt change in section assumed, this value of stress might be expected to be conservatively high. Yet the value is suffici-ently below the Mill Test Yield and ultimate values of the material (48,500 psi and 75,700 psi respectively) that a failure of the type observed would not be expected to arise from these axial stress levels. It is pointed out ; however that a third contributor to the axial bending stresses will be the thermal disconti- ' nuity stress due to bulk temperature gradient - along the nozzle and valve body walls. This gradient results from the termination of insula-tion at the valve and the greater heat dissipa-tion capability of the valve volume and surfaces. After the original stresses as described above were presented, Mr. Pulito stated that he thought the stresses due to step in thickness were unrealistically high as far as the conditions at the point of failure were concerned. Because the failure occurred at the ' transition between the Schedule #40 (.28") and Schedule #80 ( .43") sections of the nozzle,'it was felt that only this change in thickness should be considered for the " Bending stress due to step in thickness." By slide rule calculations they deter-mined that this stress (the pressure discontinuity) should be approximately 44G of the value given in the original presentation. Thus, bending stresses due to step in thickness were reduced to 44% of their original value for all three pressure cases and new totals < were substituted. These revised numbers are shown in parentheses on the chart of Exhibit-A. The derivation of the .44 factor was not explained, In any case the b g revised chart -is now inconsistent and incomplete W because we have: (using their methods) w wn, ,w_.,y . - r --- - - ~ r- _v_u ----~ =m , - - )
Pcg2 17 of 24 (W III. Discussion of Inspection Items: continued
- 1) Bending stresses due to eccentricity between the Schedulo #40 (.28") and the valve (1.5")
walls added to 2) Bending stresses due to step in thickness (pressure discontinuity) between the Schedule
#40 (.28") and Schedule #80 (.43") walls, but nissing are:
- 3) Bending stresses due to eccentricity between the Schedule #40 and Schedule #80 sections i
and 4) Bending stresses due to step in thickness between the Schedule #40 and valve wall. l Ebasco stress personnel agreed with the writer's comments that the stresses at the two changes in section could not be separated due to their axial proximity (about b") and that a nearly complete inter-action of the stress components listed above must be (Ph i s expected to exist. ; (_) ! Because the eccentricity stress is not large however, the analyses are not entirely meaning-less for comparison purposes. The original numbers can be taken to represent a harsh analy-sis or/and upper limit of secondary stresses. On the otherhand, the revised numbers could be taken as a lower non'-conservative limit thus bracketing the probable actual case. In neither case are thermal stresses considered. The writer asked Ebasco personnel whether they had considered thermal stresses. It v'as stated by i Mr. Labriola that thermal stresses were not expected l j to be high and were not calculated. The point was made by the writer that even though thermal stresses } were predicted to be low, they were present and because this was a failure analysis should be consi-dered. In this case establishing that the design met code requirements and allowable stress levels might not be enough to establish the contribution of real stresses to the failure.
i Page 18 cf 24 L III. Discussion of Inspection Items: continued - It was reported that no stress concentration factors were used in the Ebasco analyses. In view of the sharp corners apparent at the changes in section, especially at the point of failure, omission of SCF as an initiating cause of failure can be questioned. From the discussion of the subject, the writer drew the conclusion that only a Code pressure analysis had been performed in the original design. In addition, Ebasco stated that an analysis had also been made for seismic conditions. It was c1carly establiched, in response to questions by the writer and Mr. Potapovs, that the joint nozzle connection was not analyzed for the blow down case. Mr. Pulito said they were consi-dering supports for the valves, but it was not clear whether these were contemplated before the failure or after-the-fact fixes. He stated that in any case the final installation would meet a21 requirements. The ommission of the bending moment due to valve blow-off has to be considered a serious defici-oncy in the design-analysis. The jet reaction forces during blowdown are large and in the dase of this design, acting on a fairly long lever arm. If the valve popped prior to failure, the moment superposed on the pressure loads would have to be considered as a likely major contribu-tor to the failure. Conversely the fact of failure and relative stress icvels existing under steady state conditions vs. blow-off might be decisive in concluding whether the valvo popped or not. Further evidence to support either case will be essential to final conclusions. A design fix for the valve connection is being consi-dered by Ebasco and Westinghouse. They indicated that a number of different approaches were being studied
- and they were not yet ready to discuss these in detail. From the conversation, it was obvious that one modification would involve removing the short 6" Schedule #80 pipe nozzle and welding the valve )
i directlytothesteamline(andreinforcementglate) with a heavy section weld in the area of the 45 taper on the bottom of the valvo. 4 I L gg _
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[ \ w/ h Page 19 of 24 ) \ III. Discussion of Inspection Items: continued Mr. Erion stated that tney hoped to decide on the design modification within the next week. They are trying to have field work completed by June 7 Their position is that a design modification can be developed and work proceed without having conclusively learned the cause of the original failure. One cannot argue that a new design can be deve-loped without Icarning why the original failed, but from an AEC-Compliance point of view, the cause would remain important. There are many other connections of this type in this plant and possibly more in other plants designed by Ebasco or other A/E's. In any event, a revised design or any modifica-tion to the existing design should be subjected to a rigorous design-analysis. Beefing up the connection might not solve all conditions of overstress experienced and might further intro-duce new factors. It would be appropriate that f the AEC-Compliance review the new design and analysis both to confirm its adequacy and to provide a further reference for evaluation of the original failure. Mr. Siedle emphasized the need to know the cause of failure even if the fix involves a major change to the design. j He emphasized the offects on primary - systems as being of real concern as a result of this type of failure. Mr. Pulito described Ebasco's program to review all safety valve and piping connections that might be considered similar to the safety valve nozzle in the Robinson-2 plant. The following quantities of connec-tions had been identified for review: a) 24 weld connections similar to the safety valve (all with reduced thickness sections), b) 18 flanged connections, ('F ) and c) 40 other flanged, screwed or socket-welded h connections. j [ s G B m_g-man == $
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4 R.i's.l l *l Y ';gs; V f . z . e' III. Discussion of Inspection Items: continued He said they were looking at all reduced sections to . 1 4. determine if they acct design critoria. Mr. Pulito Q a o,4 n o statedthattheyworenotinaposition'todiscuss,jej'",y'~j;'; other plants at this time. .
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J b 1 r# b The writer would like to point out the evalua ; 'i [TJS'h
,"? f-tion of pipe or valve connections to meet Codei {
pressure and temperaturo critoria might not be ,,, J' f 4,,
,I goed enough to establish reliability. ,
Q ~ i. - safoty valvo connection is an excellent case,in.The failedo. G
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.* point. The joint was apparently sized for'cbv , . ,,1 i
.h A pressure and weight, and possibly scismic and ,.-i *
, zl
.j> wind loadings, but the most severo loading W
). ' . 4 tion of valve popping total ant. lysis.
was not considered in the M ,$ b :
;c , . ; f ' The codos, while providing'. , , g. iW #*' "'
cpecific rules and formulae for pressura design p g f 1 W . f? [ y 'l , generallyleavetheevaluationofimposedloadsJ.mcQfdB-to the designer and define the requirements foret" ,qy;iV;
~.h 6; 9 -
their evaluation only in general terms..It is i 5 i kl~ I
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oncumbent upon the designer to ovaluate cach ? s yh[&,pj
,_* p case individually and provido assurance that the y 4 3 p' 5
' i,wm.
- design moots its functional requirements whethert;,r V;ycp'
'[A'D t 9; the critoria therefor s3, ~
.; ,, by the applicablo codesare or not.
specifically A furtherprovided, review'- 'si f 'y N
'j ':h '7 .
of Ebasco's work in their evaluation of other . , $>vh[h.C
* +
connections should provide the AEC-Compliance ~# l ' s with the type of information necessary to.cstab.1 Qg@d C. - di' S lish a lovol of confidence in the criticallplan,O #,s,0',T,ry W t, - y , systems. *3, A
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Some discussion was centered about the fact ,that no, la J T-@n &A J' one in the group, particularly the Ebasco. personnel, 3 Qt
,./ h , ,pI y .\ with both nucicar and fossil steam plant background,,,g y tf$(d 1
t had experience or knowic4ge of'a similar plant" . Y ' g #l 9y ' s yb. failure. Mr. Siedle asked whether safety valves Su,E. ,
", y , fossil fueldd plants were similar. Ebasco personne1 Ij
- i. ,.
' explained that, while generally similar in Mmount . ? '[q.
S..J. '
)
Ni;. arrangement on steam drums or lines, because the p t'.Z M @1 J fossil plants operate at a higher temperature and /- f, [Q~, ri, M pressure the pipe connections were generally of heav]io
, 3,
- wall and/or in higher strength alloy materials.; 'Fromb Nh'f f the discussion, it could be concluded that these partik7 % Q J. y P pd $2 .
N, cular parameters of this design and its configurati6n! 4'c4 N ,$ ', . were probhbly unique to'a nuclear plant.3 - '#
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PQge 21 of 24 III. Discussion of Inspection It2ms: continued The above comments are important in categoricing the failure or assessing its probability of occurrence on basis cf other plant experience. A nozzle connection designed for higher steam pressures and temperatures would have a greater wall thickness, and/or use higher strength alloy materials. Thus, it would inherently possess a greater margin of strength to resist externally imposed loads - such as the steam discharge reaction. Conversely, when designing for comparatively low steam saturated steam pressures and temperatures in nuclear plants, relatively thin walls can be used in low alloy materials. Therefore, com-parably high reaction loads will cause a much higher stress and larger percentage of the total stress in the lighter sections. We think that this relationship might have been an important factor which resulted in the subject design being
, marginal for large discharge loads. The analyses of Attachment #4 show how the discharge forces are predominant in the total stress picture.
As the last agenda item of the meeting, Mr. Erion of Westinghouse explained their approach to reviewing the effect of the transient which resulted from the vr.1ve nozzle failure on the primary system. Ile expained that the pressure and temperature history during the transient were very well documented by facility recorder charts and data taking which was started by operating personnel soon after the start of the transient. The primary system experienced a 213 F temperature drop during a period of 63 minutes. The writer asked whether this was the actual transient experienced by all primary system components or an average coolant temperature change. Mr. Erion said that for all practical purposes this temperature cycle was representative of all parts of the system, lie said they had good natural circulation after the pumps were stopped - better than expected. He stated further that the temperature response of the system correlated well with that predicted in earlier theore-W tical computer analyses. EUshGpaascmc*AMMU@7mWNT mmAmmvemm m a mm * ^"-"M
Pag 2 22 cf 24 Cthj III. Discussion of Inspection Items: continued p - They had initiated a program to determine the offect of the transient on all major components. Designers of the vessel, pressurizer, steam generators, pumps, etc. were given the transient data and asked to respond as to its effect on their components. Mr. Erion reported that most of the groups contacted had respon-ded and no problems were identified. The report of the transient effect on the primary system referred to above was brief and more in the form of a progress report than a technical account. It is of course ^ ndesirable to experience transi-ents in components which will reduce their remaining usage, particularly early in the life of a plant. The effect of the transient on usage is not obvious nor can one categorically define its offect on all system components. Each area of high stress and thermal related usage must be g] looked at individually. It is understood that the AEC-Division of Reactor Licensing will be reviewing the applicants evaluation of the transient effect on the entire primary system. The writer.~ wishes here merely to endorse the AEC y position in assuring that the system has not suffered severe damage by making a thorough review of the reported transient effects. t o w=w-- ,-an-~-- - -
Paga 23 of 24 I IV.
References:
1 Inquiry Memorandum - Carolina Power and Light Co. (II. B. Robinson No. 2) License No. CPPR-26 Docket No. 50-261, W. C. Seidle, SRI, CO-II to J. P. O'Reilly, Chief, Inspection and Enforcement Branch, CO-Hg dated April 29, 1970 2 Inquiry Memorandum - Carolina Power and Light Co. (II. B. Robinson No. 2) License No. CPPR-26 Docket No. 50-261, W. C. Seidle, SRI, CO-II to J. P. O'Reilly, Chief, Incpection and Enforcement Branch, CO-Hg dated April 30, 1970 G S = -m ____ ,. . . - _ _ , _ _ _ -
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. . EXHIB/T %~
STRESSES FOR SAFETY VAL VE /VOZZLE .._ Q A6 pgESENTED BY EBASCO, 47 5-/S-70 MEET /N'G l CODE ALLOV/ABLE,374ESS v/3000 PS/- Y/ELQ STRENGTH 4850RhSI } MILL : L
- L/L T/ MATE STREMGTH ' 75 700 PS/ J TEST VAlud3 PRESSOQE' .
m' 900 PS/ //40 AS/ /SSG PS/ *
, PR/ MARY STRESS ..
LONG,. STRESS DUE ho PRE $50RE 4GGO 5900 ' ? 7020
,. LONG. STRESS DUE TO W6tGHT . 96 0 9GO 96 0 LONG. STRESSLDUETO JV/HD 75 0 73 0 75 0 TOTAL PR/ MARY STRESS 6370 7G/O 8730 SECONDARY STRESS: -
DUE TO-BENDING STRESS 26S0 3350 4000 ECCENTR/C/TY Duf:TO STEP /5050 /9060 !?G70 - BEND /NQ STRESS TN -7H/ctNass
~
Q (GGoo). .(8400) (19co) TO TA L. SECONDAR9 STRESS 1770 0 22 4/O 2GG')O
~
(9250) . (/3620) (/3900) PR/MARV + SECONDARY. 24070 _ 30020 3S400 STRESSE3 ;(/.SG20), (19360) ;(22G30)
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MATHEMATICAL MODEl? MARTSVfLL&;dC; . 1 l
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- DUE 70 BCCSM7R/C/7Y , PARAWlETERL INC.
u. gq. 70A - DC -G72
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- EXHISIT *B*'
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SCNEMATIC ARRAM EMEWT OF MAld STEAW1 LIME 0 SAFETY VALVES AS PRE 34H7ED SY WEST /NGHOL/SE . AT 3-)5-7C) WlEE7tHG. STSAM CTENE&iORS ~ '
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'.QENER4LCQ/LTodR OF f4/ LURE (moes ca cursme) SA/*ETV VAL VA NO2 2 L E '
do TE iFA/L ED SUUAC4 DESCRIPT/Qd - M. B. 208(WSOM do. 2 - ' AS DESCR/8ED BY wEsritailows' , HApTsv/tts A c. -
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i Pcgs 1 of 21 i !^ Rd ' 4 Stress nalysis of - Failed Main Steam Safety Valve Nozzle II. B.* Robinson No. 2 Carolina Power and Light Co. , Attachment-No. 4 To Report DC-67 w May,23, 1970 Prepared for: U.S. Atomic Energy Commission Division-of Compliance AEC Contract AT(ll-1)-1658 . Task'A ' Parameter -No. 69-70 A '
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by:- M ' Walter :J. Foley , P.S.
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W. lf -y and Lb 0 'L &
/70 Robert.S. Deah,-P.E.
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# 41s i, s - od"0-- Consulting Engineers
#"*". Elm Grove, Wisconsin
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. . . . . . . . . . . . . . . . . .. . .1.._ ........ . .. T o ..R e p o r t.. D C - 6 7.. ..... .. . . . . . . . . . -
d Table of Contents 4 Page Description
, l , Cover Sheet 2 . Table of Contents 3 Introduction
'4 Assumptions
[ 5 Conclusions
'6 Summary d.1 Description of Computer Runs 12 Description off Computer Program
- L13 Analytical.Model' 15 -Dimensions;and Elements
- O g 17 Ten- r tore oistrib. tion
- t. .
20 Stress Caused by Posaible; Popping ' 9 f ny
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. . . . . . . . . . . . . . . . . _ . . . . . . . . . . . . . . _Io..Repor.t..DC-67 .. .. .. . . . . . . . . . . . _ . . . . . . . . . . _ . . . . . .
Introduct_lon
~
The purpose ~ofLthis attachment is to make a theoretical stress amalysis'of the failed safety' valve nossle~,in i order to determine tentatively what loading conditions. existed at the. time of failure. Only the steady state hot operational condition at 900 psig and no load is-investigated. Thermal stresses are computed for four=(4) arbitrary temperature distributions which tion. are -intended to bracket !the actual thermal condi-The effect of possible.. popping is' analysed. Results are summarized in convenientIform for evaluation of.the report which will be submitted.by the licensee. Assumptions, the analytical model', assumed temperature
. C tnbutions, computer runs'and analytical methods are described. Tentative conclusions are made on the basis-
' .of incomplete information available at this time. ' f O 5 l t
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_.. ... .. . ._...Io.. Report..DC.6 7. . ~ O assumptions -
- l. The actual temperature distribution of the upper part of the safety valve nozzle at the. time of- "
failure is bracketed by the arbitrary temperature -/ distributions shown on pages 17 through 19. .j
- 2. The lower part of the safety valve inozzle.and .the steam line.were at 532F uniform temperature at the time of failure.- .o ,
~
1 3. The stress concentration factor of 1.5 is reasonable 4' for the -three (3) locations shown on page 10. y
- 4. The safety valve nozzle is described with adequate . .F accuracy by the' sketches on pages 13, 15 and 16. u-5;maressure and thermal stresses of the nozzle are not affected significantly by the- field weld backing ring.x Beca6se exact information about the backing ring' '
is not available to this writer, it is not possible ^ to present complete justification for this assumption. '
! i
' Howsver, partial justification is obtained by noting ,
that pressure stresses-are.quite low and that-both . O t8e 6'taca scaea te 40 give aad the decat a riao are at 532F for the condition analyzed. .
. - a r.
n
- 6. Pressure stresses do not' drop when the safety valve "~
pops. This assumption is justified by noting that operating .g"j,[ pressure stresses are quite' low;and that the source. of pressurized steam is very large. . 7, 1 7 The nozzle was not weakened by progressive failure.
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. To..Rcpor t DC-67J .... . .. .. . . .
s i-( Conclusions.
- 1. The graph on Page 21 indicates that pressure plus thermal stress without stress concentration is only about one-sixth as great as the minimum ultimate tensile stress. It does not appear reasonable to regard pressure plus thermal stress as a primary cause of this overload failure.
- 2. The graph on Page 21 shows that discharge thrust could have been the primary cause of the overload failure.
This graph will be useful in, evaluating the report which will be submitted by the licensee.
- 3. have caused the fracture to Discharge thrust could,Mr.
propagate according to Cunningham's explanation on pages 13 and 14 of Report No. DC-67. J I e e
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Summary
- 1. Pressure and thermal. stresses are summarized on pages 7 through 10 for locations,in the vicinity '
of the failure.
- 2. Stress caused by possible popping is calculated in general form on Page 20.
- 3. Total stress caused by internal pressure, temperature gradients and possible discharge thrust is expressed
] 7 je graphically on Pagep21. .: .,
- 4. Assumed temperature distributions'are described on Pages 17 through 19.
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u sv..M.Foloy oare. 5-24-70 sua;ccv . . . . . . . . snecT no. 11. .o . . . CH K O. DY ,', J .. . D ATE . . .. ..Attachr.cnt . No. 4. Jos no. ..g...
. . . . . . . . . . . . . . . . .._..T.o..Re po r t.. DC-67._ _ ._ . .. ..
O Description of Computer Runs _, .Run Date Description No. 1 5-21-70 DC67 at 900 psig and 70 F without SCF 2 " " " " " " " . with SCF 3 " " " " " " Temp 1 without SCF 4 " " " " " " " with SCF 5 " " " " " " Temp 2 without SCF 6 " " " " " " " with SCF 7 5-22-70 " " " " " Temp 3 without SCF 8 " " " " " " " with SCF 9 " " " " " " Temp 4 without SCF O 10 y-
;y
" " " " " " with SCF Note 1. Computer input sheets and printouts are maintained in file by Parameter, Inc.
- 2. Temperature conditions " Temp 1" through -
" Temp 4" are described on Pages #7 through'89.
4
- 3. Stress concentration factors are described.
- j. on Page iO.
4.- The computer program is described on the p following page. O
$' CY.Wa.',EC1Gy oATE 5-24-70 suaacer __
sucer no 12. or.. . .
- '< 1;cHxo. sv. . om ic .. . .. . Attachtant No.. 4 ;ou No- .
. _ . . . _ . . . _ . . . . . . . . ..To..Repor t DC-67. . .. ..
p u& .[ - Description of Computer Program 8 Finite Cylinder Stress Program A.O. Smith Corporation 2 Data Systems Division Library No. 05120 s
, This computer program is used for determining-L pressure vessel stresses in accordance with the '
l' ASME Nuclear Code. Stresses and motions are printed out for combinations of internal and external pressure, mechanical loads, and axial [ and radial temperature gradients. Metal temperatures and thermal moments taken from printouts of the Generalized Heat Flow and Thermal Moment programs are used as computer input. The
' effects of local flexibility are introduced readily at the analyst's option. Interaction' .
with bodies such as tubesheets, hemispheres and bolted closure heads is analysed by using the Simultaneous Equation Program to solve compati-p1 bility equations of two or more bodies for which v influence coefficients have been-determined.
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gc A - P (k . Analytical Model - continued x. @g s, m?Because of:a characteristic of the computer program, the y,' ' nos' ale is divided into two bodies as shown on the $r ; . .
- Preceding page. The upper part of the nozzle and'the gf point'of failure are included'in Body 1,-which is_sub- *-
y;po l . divided-into ten.(10) short cylinders. The steam line,
, z thelreinforcing pad, the: fillet welds-and a short length g--- .of;the?noskle'are modeled'as Body 3, which is subdivided y -
, g intoisix'(6)Jahort cylinders. In order to-seal internal
, ; __. .a pressuref a very thin " infinite" cylinder is located at
.Q?: , each end of the; analytical model. y, . To; avoid. complications, the upper part of Body 1 is c .modeled.without steps. This simplication is justified
; s by'the practical observation that the flanges and the ;[
counterbore neck significantly. do not affect the failed Schedule 40 nozzle py, b E
- . The complex transition from nozzle to steam line is 1
' modeled;very simply as_a reasonably thick cylinder on F 2 7thetsame axis as the nozzle. This simplification is
~
' M.
-justified.by noting that the upper fillet weld'is approximately a."long cylinder' length".fron. the point of.
M failure.- ;In other words, 'neither the steam 'line nor the imaginary?short" cylinders. interact'significantly.with-
-the failed Schedute.~40 nosale neck.-
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