ML20151Y285
| ML20151Y285 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 04/27/1988 |
| From: | Cockfield D PORTLAND GENERAL ELECTRIC CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| RTR-NUREG-0691, RTR-NUREG-691 NUDOCS 8805040412 | |
| Download: ML20151Y285 (63) | |
Text
{{#Wiki_filter:PortlanciW ElectricCotaxwly David W. Cockfield Vice President, Nuclear l April 27, 1988 Trojan Nuclear Plant Docket 50-344 Licenso NPF-1 U.S. Nuclear Regulatory Comission ATTN: Document Control Desk Washington DC 20555
Dear Sirs:
Examination of Feodwater,,p.educers - Trojan Nuclear Plant During the 1987 refueling outage at the Trojan Nuclear Plant, feedwater piping reducers into all four steam generators woro replaced. Reducers "A",
"B", and "D" had exhibited flaw indications by ultrasonic testing, while "C" reducer showed no indications. The reducers woro divided amongst Portland General Electrle Company (PCE), the Nuclear Regulatory Comission (NBC), and Electrle Power Research Institute (EPRI) for examination of the flaw indications. Reducer "C" was provided to EPRI for uso in an experi-mental loop. Attachment 1 is a report summarizing the findings from destructivo examination of the "A" and "B" reducers performed by PCE.
PGE also evaluated recommetdations made in the NRC study on foodwater line cracking (NUREG-0691, "Investigatlon and Evaluation of Cracking Incidents in Piping in Pressurized Water Reactors") to determine thele applicability to problems experienced at Trojaa. Our review found that the recommenda-tions applicable to operating plants are valid for Trojan. PCE plans to perform augtr. anted inspections of the reducers as recommended, with accep-tance criteria for indications conservatively grounded in the results of the above-mentioned destructivo examinations. Attachment 2 describos our present plans for the inspection of the reducers. Other feedwater piping experiencing thermal stratification will not receive augmented inspections. This piping will have all been replaced in the 198/ and 1988 outages and will experience no greater time of exposure than the new reducers. Since the more severo geometry of the reducers causes them to be the preferred location for cracking, their inopoetion and conservativo evaluation will provido advanco warning of potential problems with the feedwater piping. 8805040412 080427 f PDR ADOCK 05000344 - G PDR pig e sw fron swa rtxwu ong:n 9 coa
l Pordarid Generad Electric Corquiriy Document Control Desk - April 27, 1988 Tage 2 e' Long-term remedial measures addressing the cracking of'the feedwater reducers are currently being investigated. Sincerely, e
} W4 ,
gf/' 'l Attachments t c: Mr. John B. Martin Regional Administrator, Region V U.S. Nuclear Regulatory Commission Mr. William Dixon State of Oregon Department of Energy Mr. R. C. Bare *' NRC Resident Inspector : Trojan Nuclear Plant Dr. L. C. Shao, Director Division of Engineering and Systems Technology , ; i U.S. Nuclear Regulatory Commission _, Mr. Larry Becker/Mr. Frank Ammirata (2 copies) Electric Power Research Institute Non-Destructive Examination Center i l I a 1 I
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() e , i .[, ! Trojan Nuclear Plant Docunwnt Control' Desk i Docket 50-344 Attachment 1 . License NPF-1 l Apell 27, 1988 jl Page 1 of 37 i EXAMINATION OF REDUCERS REMOVED , 1 f FROM "A" AND "B" STRAM GRNERATORS ,
'l i Summary 4
j The destructive examination of the reducers removed from the "A" and "B" . steam generator feedwater piping revealed that. internal diameter initi-4 ated circumferential cracking existed at the counterbore location, approximately one-half inch from the centerline of the "V" groove: reducer 4 to nozzle welds (Figures 1 and 2). The maximum depth of the cracking'was 4 ' O.533 inches for the "A" reducer and 0.324 inches for the "B" recucer. In._each case, the maximum crack depth was located near the top of the ! reducer. Outside of this lobe of maximum penetration, the average crack : l- depths were about 0.075 inches and 0.050 inches, respectively. All ; appearances suggest that while most of the-crack was not propagating,_the lobes of maximum penetration constituted active crack. fronts. There is no reason to believe that this' process would.not have led to a through ; wall crack and subsequant leakage. However, the probability of cata- ! strophic failure is judged to have been low, with the most likely . scenario being a 5- to 6-inch-long breakthrough with a predicted leak rate of about 10 gpm. ,
- Background i
i The documentation dealing with this problem since its recognit. ion in 1979 is voluminous. In addition to various internal memoranda discussing non-l destructive examination results, etc. there have been two analyses , , performed. 1 j These were:
\
- 1. 'ACAP 9613 "Integrity Assessment of Feedwater Line Indications -
Trojan Nuclear Plant", September 1979, 9
- 2. Letter POR-87-568 transmitting, "Fracture Mechanics Assessment",
June 18, 1987 (this reference is included in Attachment 2).
- The material used in fabricating these reducers is American Society of '
Mechanical Engineers (ASME) SA 106, Grade B, medium carbon steel intended for high-temparature usage, with minimum specified yield and tensile
! strengths of 35 and 60 ksi, respectively. The Ladish test reports for the material show the following chemical cnd mechanical properties.
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c Trojan Nuclear Plant Document Control Desk I Docket 50-344 Attachment 1 , License NPI'c 1 April 27, 1988 ; Page 2 of 37 i j . Chemical Analysis R Car'on 0.28 w/o Manganese 1.05 w/o Phosphorous 0.023 w/o Sulfur 0.035 w/o i Silicon 0.31 w/c ; t Mechanical Properties l Yield Strength 53.2 ksl Ultimate Tenslie Strength 88.4 kai ; Percent Elongation 29.0 ' Percent Reduction in Area 59.0 Charpy Impact Properties. 0*F Energy Absorbed % Shear Lateral Expansion 23 ft-lbs 5 0.019 inch 41 ft-lbs 17 0.035 inch 50 ft-lbs 29 0.045 inch 1 w i The chemical analyses conducted on the removed "A" and "B". reducers (Figure 3) are close to the ladle analysis and provide further verifica- . ] tion that the material is within specification and would be above its l ductile to brittle transition temperature during operation. ' The generally advanced mechanism for the cracking seen here is low-cycle. J' high-stress fatigue or corrosion fatigue due to stresses caused by thermal stratification. In many cases, the cracks stifle at about 0.10 inches. This implies that the thermal stresses are primarily located in the inner 3 bore of the pipe. In some cases, extensive crack propagation occurred in i
. a relatively short period of time. This inplies substantial through-wall !
( stresses. Trojan appears to be unique in that the cracking is extensive, j yet apparently has occurred over a relatively long period of time. 1 Examination Results It was decided during the 1987 Outage that all four reducers would be replaced. Reducers "A", "B", and "D" exhibited flaw indications by I ultrasonic testing (UT), while the "C" reducer was clean. The reducers l were divided up for investigation purposes as followed:
"A" Reducer - A 5.6-inch-wide 8-inch-long section was removed from the 90-degree position (as viewed from the steam generator) and pro- l
! vided to the Electrical Power Research Institute (EPRI) for UT ! l investigation purposes. The rest of this reducer was examinea by PCE. I 1 l
Trojan Nuclear Plant Document Control Desk Docket 50-344 Attachment 1 . License NPF-1 April 27, 1988 Page 3 of 37 ! , "B" Reducer - A 5-inch-wide x 8-inch-long section was removed for - EPRI. The rett.was examined by PGE.
"C" Reducer - %is was provided to EPRI for use in an experimental loop. l "D" Reducer - This reducer was given to the Nuclear Regulatory; Commission.
A typical reducer ("B") is shown in Figure 4. The two cuts made to remove the EPRI sample and the counterbore area are indicated as 1 and 2 in Figure 5. A typical internal view of a reducer ("A") is found in Figure 6, with the EPRI section indicated in Figure 7. Note that all angular orientations i discussed in this memorandum are based on viewing the large end of the reducer from the steam generator and proceeding in a clockwise direction. This is opposite to the convention employed in some of the historical documentation. Therefore, any reference to angular position in such' documentation should bo interpreted in accordance with the convention i described in the individual document. l The cracks varied in appearance and number,e 'spacially where cleanup for liquid penetrant examination had taken place. This can be seen in ! Figure 8 ("A" reducer). As-found cracks are shown in Figures 9-12. . Extensive multiple cracking was most prevalent in the'"D" reducer , (Figure 12). j The ring sections cut from the "A" and "B" reducers are shown in 9 Figures 13 and 14. The EPRI section included the angular portion from j 70 degrees to 110 degrees. The numbered blocks were then cut from the { d rings, ground to reveal the local cracking, and measured in some detail. ' These measurements were nonnalized to compensate for material lost in cutting, and the flaw depth and wall thickness tabulated as a function of j both angular position and linear distance along the outside circumference. l These results are plotted in Figures 15 and 16. ! The presence of the major lobe of each reducer crack in the 0-degree position is quite evident from these figures. Also evident ara mtaller lobes at about 115 degrees, 200 degrees, and 285 degrees in the "A" reducer and at about 280 degreco in the "B" reducer. These features j have been noted in other plants, as discussed in the letter of June 18, i 1987 (MT-SME-6056). lI Active cracks have been observed to be relatively clean with sharp tips, while inactive cracks are best characterized as being blunt, wide, and full of oxides. Where deep active cracks are found, they are usually unitary, as contrasted to the multiple cracking associated with the l l
Trojan Nuclear plant D(cument Control Desk Docket 50-344 Attachment 1 Licenso NPF-1 April 27, 1988 pago 4 of 37 smaller inactive cracks. This een be soon in the following examples of opened cracks and metallographic sections from both the "A" and "B" reducers.
- 1. Figures 17-18 and 19-20 are pairs showing the opened fracture sur-faces at about 5 desreca and 25 degrees from the "A" reducer. The fracture surface in the second figure in each pair has been chemi-cally cleaned in an ultrasonic cleanor to remove some of the oxides.
It is interesting to note that the first 25-30 percent of the crack surface is more oxidized, more pitted, and much more heavily deposited with coppor than the remainder of'the crack. Measurements made through a stereoscopic microscope found 13 clear crack arrest marks on the five-degroo fracture surface. The loca-tions are tabulated and plotted in Figure 21. It is not clear what meaning can be ascribed to the 13 crack advance events. Are they years, periods of hot standby, etc?
- 2. Figure 22 shows a complete picture of a transvorso section through the "A" reducer at about the 360-degroo point. Note that the crack is oxidized near the top, but is generally direct and sharp pointed.
- 3. Figure 23 is a similar section taken at the 115-degree point, which is the peak of one of the smaller lobos of penetration. It can be argued that this lobo is not moving in a regular fashion, because of the heavily oxidized condition at arrest points (Figure 24) and at the cracx tip (Figure 25).
- 4. Figure 26 is a transverse view of the insido edge of the pipe at or near the lower penetration region in the 210-degree position, where i rultiple cracking was found. A profile of the longest crack emanat- l ing from this edge is in Figure 27, and the oxidized crack tip in l Figure 28. This region appears to bo inactivo.
- 5. Figures 29 and 30 show the last small lobe at 285 degrees. There is some evidence of progressivo crack advance in the cleaned specimen 1 (Pigure 29).
- 6. /igures 31-34 show the type of cracking experienced on the "B" reducer. The "B" reducer cracking was only about 60 percent as '
deep at maximum as that in the "A" reducer. It also had a greater i tendency to consist of shallow multiple cracks. The counterbore i transition on the "B" reducer was somewhat smoother than on the "A" reducer, which had a shallow machining gouge at this lodation. The ef fect of the abrupt transition was to make it easier to s+ set a singlo crack in the "A" reducer and propagato it in both directions. Its effect as a stress raiser was limited to the early stagos of cracking, since a crack of more than a few thousandths of an inch l
v l
- Trojan Nuclear Plant -Document Control Desk Docket 50-344 Attachment 1
- . License NPF-1 April 27,'1988 page 5 of 37 i
, i
- deep is a auch greater source of ctress amplification. The lack of !
such a stress raiser resulted in the "B" reducer _ cracking exhibiting multiple origins (tear ridges), as may be seen in paired Figures 31-32 (350 degrees) and Figures 33-34 (50 degrees). The ridges parallel to the direction of-crack propagation are boundaries between separately initiated cracks, which have'not yet come together on a common plane. However, even though the "B" reducer cracking had multiple origins for the deepest'part of the crack, there was only ; one crack ~in this most highly stressed region. i 7. Figure 35, a metallographic section through the cracking at about the 10-degree position, shows a relatively sharp and straight crack.- Although oxidation is present in the crack, especially at the arrest , points (Figures 36-37), the crack is still fairly sharp'(Figure 38). The above should be contrasted with the crack morphology at about I 125 degrees. Here the crack is heavily oxidized and has branched at i
- the end (Figures 39-40). This region is probably not propagating. _
- 8. Figures 41 and 42 show multiple cracking at-about 190 degrees. ;
Figure 42 is a closer view of the second crack from the right in ; Figure 41. Note the oxido fingers extending perpendicular to the ! crack about halfway down Figure 42. This presumably is a crack ' arrest location. j j j 9. Figure 43 shows an area where nonpropagating multiple cracking l 1 existed. As usual, one crack was deeper than the rest. When
- deformed to expose this crack at the 261-degree location, the }
adjacent cracks opened up (Figure 44). , 1 Discussion
.)
Stress analyses conducted on "A" reducer were done in 1979 and 1987. The i first (WCAp-9613) provided a fatigue crack growth analysis, and a criti- I , cel fikw size determination. Assuming a 0.100-inches deep circumferential l crack, a growth to 0.1006 inches was predicted in 10 years. The critical l flaw size for the worst-case condittors [ thermal + deadweight & operating basis earthquake (OBE) loads] was found to be a aontinuous ,. circumferential crack at the inside sue.~ece with a depth of 0.510 inches. i 4 The crack depth in the "A" reducer was b6 'ieved to be about 0.100 inches. Although UT measurements in 1980 suggestes a depth of 0.29 inches and
- radiographic triangulation measurements in 1981 yielded a figu.0 of 0.28 inches, the majority of measurements (UT) through the years indi-
^ I cated that 0.100 inches was a tasonable value. 1 1
~
This was still the situation in 1987 when an allowable flaw depth ! analysis and a crack-growth rato analysis were performed. Using somewhat ' 1 different methodology for the former analysic this time, a worst-case 1 I 3
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Trojan Nuclear Plant Document Control Desk Docket 50-344 Attachment 1 Licenso NPF-1 April 27, 1988 Pago 6 of 37 result of ag/t = 0.396 was arrived at as an acceptable flaw depth for normal / upset / test conditions, and a value of 0.41 for emergency / faulted conditions. These results correspond to a depth of about 0.34 inches for a flaw length to circumference ratio of 0.3. The analysis is based on ASME Section XI, Task Group Hooting Minutos of January 19, 1987 and EPRI Final Report, Project RPl757-51, October 1985. The fatigue crack growth analysis presented earlier did not include , thermal stratification concerns. After adding thoso stresses to the ! evaluation, it was concluded that reasonably conservative comparison with other plants indicated that Trojan could operato for 6,500 h u rs at hot standby before exceeding the previously calculated limits (POR-87-568). l BC, sed on Plant historical data, this implied a reducer life expectancy of over 27 years. l The prediction of crack growth rate and anticipated crack shape is dif-ficult. It cannot be said with certainty that the observed situation l would lead to a through-wall crack. However, consideration of the fol-I lowing qualitative analysis suggests that these cracks would be likely I to continue to propagato. The analysis is based on the stress-intensity l factor solutions for semielliptical surface cracks in a plate of finite dimensions, according to Raju and Newman. The geometry of the situation is shown in Figure 45. The stress-intensity factor "K" is, of courso, strictly applicable to linear clastic materials only. However, the rela-tivo magnitude of the crack-driving force at different times may be approximated. Calculations indicate that the crack-driving forco rises by a factor of threo as the crack depth increases from 0.17 inches to 0.68 inches. Unless the applied stri,ss, 6, drops more rapidly than this, a propagating crack would be expected to continue to advance. t The solution also suggests that a somielliptical crack will tend to becomo semicircular. The eventual crack breakthrough should thus be on a maximum 6-inch front. Since the calculated critical through-wall flaw is 23.5 inches, the risk of catastrophic failuro is minimal. Data also indicates that the expected leak rate should be around 10 gpm. I New Reducers l Figure 46 shows the shape which was recommended for the new reducer l counterbore. It differs from the old one, primarily in that the ; l transition is moved from about 0.5 inches from the weld centerline to ; j about 1.75 inches from the we'd centerline. Improvement of the surfaco ; finish was also recommended. An error in machining the new reducers resulted in the counterbore i transition being located again at 0.5 inch from the wold centerline for I "B", "C", and "D" reducers. The "A" reducer counterbore transition was l machined at 1.7 inches from the wold centerline. JWC/2393P , 1 I !
_ ~. Trojan Nucisar P1&nt Docunent Control Dssk j, Dockst 50-344 Attachmsnt 1 Licenss NPF-1 April 2 7, 1958 Page 7 of 37 1 /- WP No. 2 Qy 14" Elbow Steam Sch.60 16" Feedwater 16x14 / Generator Nozzle Red. Y [ WP - Weld Preparation
-WP No.1 1
Figure 1 Nozzle to Feedwater Piping ; Feedwater Nozzle Weld Reducer Y
-,43-1 1
0.75" Pipe Wall l l J O V l a
/ l
' h ! 150Max. ! Location of incication Observed i i Reducer O.D. = 16.16 in. Figure 2 Nozzle to Reducer Weld I -
. - . . . . . . . . - . ~ - . . ,
t- Trojcn Nuclear Plent Docunant Control Desk Docket 50-344- _ .
. . . . - ~ Attachment 1- -
, Licensa NPF-l' April 27, 1980 '
. SC$ N. Lagoon Acnve Page 8 of 38 co=otevetro= i=seection aca ctstawetive toen=e
- atseiasa insesetion P.O. Box 17126 . a s p . c a .,,,,c . t.o,,
3 casuicas amat,sie 4 18"'a* Port!and, Oregon 97217 0126 s'=,sicas esetime assa. =e i Pnone. (503) 2891778 i ! September 17, 1987 ' Oregon Analytical Lab c/o Portland General Ele,tric Co. , 14655 S.W. Old Scholls Iarry Road Figure 3 Beaverton, Oregon 37005 Attention: Mr. Jeff Carter Subiect: Chemical analysis performed on two (2) steel samples submitted on 9/14/87, per your PO Number 8691. REPORT: Item Steel Samples #6A and 6B Chemical Composition:
# 6.*. #6B l Carbon (C)% ....................... 0.26 0.27 Mang ane se ( Mn ) % . . . . . . . . . . . . . . . . . . . . 1.00 1.02 ~
i Silicon (S1)%...................... 0.36 0.36 , Chromium (Cr)%..................... 0.21 0.22 : Copper (Ct,)%....................... 0.17 0.18 Nickel (Ni)%....................... 0.11 0.11 j Molybdenum (Mo)%................... 0.05 0.05 Vanadium (7)%...................... 0.008 0.008 Al um i n um ( A1 ) % . . . . . . . . . . . . . . . . . . . . . 0.012 0.013 ; 4 Phosphorus (P)%.................... 0.011 0.012 4 Sulfur (S)%........................ 0.033 0.035 i I 4 Respectfully, ; NORTHWEST TSSTING LABOPATORIES, INC. j i . d0 _ j John Kral, 2alys I 1
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A & L ___
- Howard Holmes, i Assistant Supervisor, Chemistry Report Number
- 308637 l
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~ , , ..o ; c ..o., .. .. :: 9 ,n . ...p. ,,; +, .. ,. .- .. .. s -~ N Y,M ' ' .I .,4 . ' . '{:_ ,; " . ;.* * -..f " l. ~ . l , ...c' - = s.Q:$ .v .. n , . & h- .Qie shi f$7 ?h' C' ~. ^ . L i .' I ';~ T . k Y Y f. A m.e...v.< m. m m : m. . . p[., , 9.%,f?'f;.f.?.i..o..-l i ' G l: Figare 44 ' I. .L . . f- ~} ,' R e ri , er ?.(i , ,1 m e :{ ' 1 ] + ; ;: ] ,, c;, g( p. ) g g (g, ;> g } -.'. jr ) .f n 4 ~- ~ _ _ _ - '----- __ma' ,A A M ~ t g Trojan Nuclear Plant Document Control Desk Docket 50-344 Attachn. ant 1 License NPF-1 April 27, 1988 Page 36 of 37 l l l 0 I I I I K = CoVtw/& - I' I C l a/B /q s ! % alc 9 0.2 0. 4 0.6 0.8 ' I ~ O.617 0.72C 0.899 1.190 s s O' ~ ' I % 0. 2 45' O.990 1.122 1.384 1.657 ~ ' 6 90* 1.173 1.3S9 1.642 1.851 I ,' \ O' O.767 0.896 1.080 1.318 l s 9 I 2c 04 45* o.998 1,075 1.247 1.374 hs 90* 1.138 1.225 1.370 1.447 / % % 0* 0.916 1.015 1.172 1.353 ( ; %, 06 45" 1.024 1.062 1.182 1.243 's 90* 1.110 1.14S 1.230 1.264 0* 1.174 1.229 1.3SS 1.464 1.0 45* 1.067 1.104 1.181 1.193 < 90* 1.049 1.062 1.107 1.112 l V P G W 8 1 l i Figure 45 Stress Intens1ty Factor (K) I ) l Trojan Nuclear Plant Document Control Desk Docket 50-344 Attachment 1 License NPF-1 April 27, 1988 Page 31 of 37 , l i Figure 46 l l RECOMMENDED WELD END PREP FOR STEAM GENERATOR FEEDWATER NOZZLE TO REDUCER l l I I n \ / \ WELD v \J I _ _ 4 TD I 2T M lh) _ IO* I:A) R O U T WrrH HAMD 4 E.ND W.* R ADius FILLET
- AM FoR. ~ 179 RM6 FIN 16N ON Cout4T"EltBoRE ( TRAN6tT~loN
[RMS - Root M EAN sq u. ARC,] . Unobstructed access of 2" minimum each side of weld. OD reinforcement 0.030" is okay, but weld crown must blend ! sm othly and gradually with base metal. 3:1 max. transition. i (If steeper transition is necessary, or if unobstructed access ! of 2 inches is available from one side only, then grind weld I flush with base metal). Scribe line 4 inches from center of root opening prior to welding so that true center of weld can be established during UT. < j y-Trojan Nuclear Plant Document Control Desk Docket 50-344 Attachment 2 License NPF-1 April 27, 1988 Page 1 of 24 NONDESTRUCTIVE EXAMINATION OF STEAM GENERATOR FEEDWATER INLET N0ZZLE 16- x 14-INCH CONCENTRIC REDUCERS INSTALLED DURING 1987 OUTAGE
Background
The internal circumferential cracking of these reducers was identified in the "A" reducer in 1979 and in the "B" and "D" reducers in 1985. After replacement in 1987, the "A" and "B" reducers were destructively examined. It was found that while the examination. process had detected the lineal extent of cracking fairly w'11, the actital crack depth-(0.533 inch maxi-mum in "A" and 0.324 inch maximum in "B") substantially exceeded the measured depth of about 0.100 inch (see Attachment 1). This discrepancy is apparently due to the difficulties inherent in the geometry and material changes caused by the location of the counterbore transition at 0.5 inch from the centerline of the weld. The cause of the cracking is thought to be thermal stratification which causes thermal fatigue of the reducer at the transition. Absent such stratification, there is no reason to expect such cracking, present Situation The present configuration of the new reducers is, except for the "A" generator, fundamentally the same as before. The counterbore transition ' for the "A" reducer is 1.7 inches from the weld conterline, while the ~ other three remain at 0.5 inch. It was intended that all four reducers have the transition at the 1.7-inch distance. However, a machining error ) precluded this. Other improvements that were incorporated included a ' transition radius of 0.5 inch, a taper of 10 degrees, and a greatly improved internal surface finish. Since these improvements were thought to be of more significance in improving fatigue performance than the l location of the counterbore transition, operation in a use-as-is ' condition was approved. The 1.7-inch distanco from the weld centerline was primarily intended to enhance nondestructive examination (NDE) ' capabilities. Thermal stratification continues to occur when auxiliary feedwater is i injected. Delta T values of over 350'F have been measured between the top and bottom of the nozzle. Thus, the driving force for thermal i fatigue is still present. ' Recommendations Under unfavorable conditions, such as extended periods of hot standby, extensive cracking has occurred in one reacto'r cycle (D.C. Cook 2). Because of the fundsmentally unchanged nature of our installation, it is t
.v - - . . __ _ _ ___ _ _ _ _ _ _
*- Trojan Nuclear Plant Document Control Desk Docket 50-344 Attachment 2 License NPF-1 April 27,'1988 Page 2 of 24 recommended that NDE examination of all four reducers be conducted in 1988 and 1989. If no indications are observed, this frequency could be l
relaxed /Lo once every two years. The more frequent inspection at first ' is intended to ensure.that unrecognized variables, primarily materials
;related, have not changed the previously observed rate of crack propaga-tion when cracks are present. Inspection of all four reducers is war-ranted by the varied behavior of the first four in this regard.
Acceptance criteria for indications found during such inspections will differ for the two reducer geometries. The evaluation of the removed reducers cast considerable doubt on the depth measurement accuracy where the counterbore transition is 0.5 inch from the centerline of the weld. Therefore, the conservative position for.the "B", "C", and "D" reducers is to set.the acceptance criteria at the flaw length that would extecpo-late to the critical flaw length prior to the next inspection. According to the attached analysis. transmitted by POR-87-568 of June 23, 1987, the critical flaw length is 23.5 inches. Unless there is some reason to believe that cracking in the "A" reducer (counterbore transition at 1.7 inches from the weld conterline) cannot bo adequately characterized as to length and depth, flaw size acceptance ! criteria will be set in accordance with the Westinghouse analysis (unless or until ASME Section XI criteria are applicable). 1 2 I 2 2393P l l yv1- - .__. - 7- ,c _ ---.aq-.p-... --,s.y9y.,g-e- -- wr- e yy
-' Trojan Nuclear Plant Document Control Desk l
Docket 50-344 Attachment 2 l License NPF-1 ' April 2 7, 1988 l Page 3 of 24 j 1 i Westinghouse Power Systems @,5;53 Electric Corporaticn Scs 355 l Pms:Atg't Pevsyh31315230 C3.. POR-87-568 June 23, 1987 l Mr. A. N. Roller, Manager Ref: W Letter Nuclear Plant Engineering Department POR-87-564 Portland General Electric Company Dated 6/12/87 121 SW Salmon Street l Portland, OR 97204 l Attention: Mr. C. P. Yundt - l Portland General Electric Company l Trojan Nuclear Plant Steam Generator Feedwater Line Crackino Evaluation Revision !
Dear Mr. Roller:
At your request, we have revised our report on the fracture mechanics l assessment of the Trojan feedwater line indication previously transmitted l to you via the referenced letter and is attached to this letter for your review and use. In order to expedite its delivery, this rivised report was originally telecopied on June 18th to Mr. Paul Yundt. Based on the plant operation data supplied on June 17, 1987 by Frank Rogan, we have included a series of comparisions between Trojan and other plants for which extensive temperature measurements were carried out. Based on these comparisons, it is concluded that at least 6,500 hours of hot standby operation should be allowed before the existing indication would propogate to the allowable limit depth. If you have any questions pertaining to this report, please do not hesitate to call us. Sincerely, T.DI L. E. Elder, Manager - Western Area U.S. Nuclear Projects JSM/mbs i
>>os ii,
Trojan Nuclear Plant i Document Control Desk j s' Docket 50-344 ' At t ac hment 2 License NPF-1 April 27, 1988 Page 4 of 24 POR-87-568 Mr. A. N. Roller June 23, 1987 cc: *A. N. Roller,JL C. A. Olmstead, IL j C. P. Yundt, IL, IA ' T. D. Walt, IL 4 4EEEEEEEEEE3$E3393REEJ H. H. Hicks, H, IL l C. J. Yarbrough, H, IL ;
- NOTE: All copies sent to Mr. Roller except C. A. Olmstead, W. H. Hicks and C. J. Yarbrough i
ties iis
. . . . . . ~ . . . - - . - - - , . ~ . . . , . . .
Trojan Nuclear Plant Document Control Desk Docket 50-344 Attachment 2
,' License NPF-1 April 27, 1988 Page 5 of 24 June 18,1987 W. H. Bamford J. F. Enrietto J. L. Hartman A. L. Thunnan Y. S. Lee
- 1. Introduction A fracture mechanics assessment has been carried out to detennine the acceptability of the indications to the rules of Section XI. There are no rules presently published in Section II for evaluation of indications in ferritic piping, so use has been made of the draft rules for acceptance of indications, as contained in reference 1. The draft technical basis for evaluation of flaws in carbon steel pipes was contained in reference 2, and follows the same general approach as the evaluation procedure for flaws in austenitic pipes (3).
The analysis of the feedwater lines for the Trojan plant was based en the actual piping loads and operational transients documented in reference 4. This analysis was updated to include the effect of thennel stratification cycles, using the extensive work reported in reference 5 Waterhanner loadings were also included in the work.
- 2. Deterrination of illowable End of Evaluation Period Flaw Derth The general approach used in evaluation of flaws in Section II is to calculate the critical timw size for the location of interest, for both normal / upset / test conditions and emergency / faulted conditions, and then to reduce these flaw sizes by safety factors to reach allowable flaw sizes. The final task of the evaluation is to ensure that the indication does not grow to exceed'the allowable size during the period of service until the next inspection. This
. _z -- .-.- --.- - .. . . - _ . . - _ . _ _ - . - . - . - _ _ - _ - . - - _ - .--
, Trojan Nuclear Plant Document Control Desk Docket 50-344 Attachment 2 l
, License NPF-1 April 27, 1988 I Page 6 of 24 same approach is used for vessels and piping, but the calculational methods may be different as well as the mechanism of crack growth which must be considered. '
, . 1 l The flaw evaluation procedure for ferritic piping sich is now available in I draft form Il3 will be used. The approach is generally the same as that published in reference 3 for austenitic piping, as are the safety factors. The j allowable flaw size is found frce tables, once the cabined primary membrane and bending stresses are known. These values were found from the loadings specified in reference 4. The value used for normal and upset loadings included internal pressure plus thertal, deadweight and operating basis earthquake (OBE) loadings, for the loop with the highest loading in l reference 4. For emergency ano tauAteo loaoings, the OBE was replaced with a f design basis earthquake (DBE). The values determined f'rca the analysis were as follows. l P,4b 19 ksi (Normal / Upset / Test) l l P ,+Pb = 36.0 ksi (Emergency / Faulted) In order to enter the allowable flaw depth charts, which are presentai as i Tables 1 and 2 the value of Sm for the piping sust be known. Frce Appendix 1 of Section III E03 the value of Se in the temperature range of operation I (200-400F) was found to be 20 kai. Therefore the ratios to be used in entering the charts were calculated to be (P,+P )/b Sam 0 96 for normal / upset conditions, and (P,+Pb )/Sa = 1.8 f r emergency / faulted coMitions. For the indications in the Trojan feedwater lines, the ratio of flaw length to circicference is 0 3, so the allowable flaw depths were found to be i af /t = 0 396 (Normal / Upset / Test) (Emergency / Faulted) a/t 0.41 The next step is to determine the fatigue crack growth for the existing indications. _ - _ __ _ _ ._- _ - _1. J.Znnz. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ -
.- Trojan Nuclear Plant , Document Control Desk Docket 50-344 Attachment 2 License NPF-1 April 27, 1988
- Page 7 of 24 3 Fatirue Crad Growth Analv3D To assess the potential for growth of the indications in the feedwater lines, all the known loadings must be taken into account. A fatigue crack growth analysis was carried out snd reported in reference 4, tich included all the design loadings. Results of this analysis are reproduced in Table 3, and 4
showed negligible growth. Subsequently an extensive investigation of feedwater line cracking was conducted, and it was found that themal stratification was occurring. This behavior was not included in the design transients, and therefore sust be considered here to- completeness. Measurements were made of actual stratified flow at five plants, during hot standby operation een stratified flow occurs. The individual plant temperature profiles were combined with flow model tests to generate number of "typical" temperature profiles, corresponding to different levels of the hot <:old water interface. These profiles were then particularized for individual plants to develop a set of transients dich would be expected to occur during hot standby operation, and these transients were applied with fatigue crack growth analysis to predict the crack growth behavior in the few plants where significant crack growth occurred and the pipes were sectioned. An analysis was also carried cut to predict the crack growth in a plant known not to have significant flaw growth, { The results of the crack growth analyses were in good qualitative agreenent with the actual crack profiles of the sectioned pipes. The results frcan l reference 5 are reproduced in Figures 2 and 3 The crack growth results were ! presented at 20 locations around the pipe circumference, as shown in the figures. These crack growth results were used to estiente the crack growth in ! t the Trojan plant, since specific measureennts of feedwater stratification are , not available. , To develop an estimate of the crack growth for the Trojan plant, the results for each of the crack growth locations were plotted as a function of hours at i i
-- em - ----,- - - .--. .- ,.,, 1 l T' l ,, lQ_ , __ , ; __ _ _ _
*krojanNuclearPlant Document Control Desk Docket 50-344 Attachment 2 1 License NPF-1 April 27, 1988 Page 8 of 24 hot standby for each plant. The hours at hot standby used here are consistent ,
with those used for thee analysis presented in Reference 5 Based on the l assumption that operational hot standby events were identical to those observed een the stratification data was collected, the total number of hours at hot standby was equal to the time at hot standby during the test period times the number of hot standby events times an uncertainty factor of 1.8. Figure 4 shows the results of these plots, for all the five plants studied, using an initial flaw depth of 0.100 inches. Figure 4 can new be used to estiante the crack growth tich scy be allowed, and from that the allowable hours at hot standby can be determined. The allowable j depth was calculated to be 39.6 percent or the wall, and adding this depth to the figure the following conclusions could be reached regarding the allowable hours at hot standby. l If Trojan operated ir. th mode of the most severe of the thennel ) stratifications, Plan' . . it would be allowed 800 hours of hot standby operation before reaching the allowable limiting depth. Operating in the sede f
- of Plant D would give an allowable 2150 hours, tile operating in the mode of Plant C would result in 6500 hours allowable. Operation in the mode of !
Plant A or Plant E, and most other plants would not result in any significant crack growth, and therefore the allowable maber of hours would be indefinite. This is because the stratification cycles and the severity of those cycles unre not sufficient to cause any measureable growth. Most of the plants dich j experienced feedwater line cracking, as found in the inspections of 1979 and 1980, fall into this category. A comparison has been made between the operating history of Trojan and other plants dich had provided data as part of a feedwater line cracking investigation. Table 4 presents the caparison based on reported data fra each of the plants. The key ccanparisons are total hours at hot standby, hot standby hours per year or operation, effective full power years, and hot standby hours per effective full power year. In all four ostegories, Trojan falls between Plant A and Plant C. Table 5 presents similar ocuparisons based on analyses wiich used site test data to determine the total hours at hot
- , - _--_- _ L = _ = = :: : = = : - . . _ _ _ _ . - . - - _ _ - _ - . ~ - . - . . - -
l .
. Trojan Nucicar Plant Document Control Desk Docket 50-344 Attachment 2 ;
License NPF-1
.- April 27, 1988 Page 9 of 24 standby. Although not bracketed by Plants A and C, Trojan is closest to those two plants in effective full power years, hours at hot standby per IFPY, and hot standby hours per year in operation.
Based on the above comparisons, Trojan has been operated most like Plants A and C. Taking the more conservative of the two from a crack growth standpoint, the operation of Trojan can bt 'ce: pared most directly to that of Plant C. Thus, l Trojan abould be able to operate for 6500 hours at hot standby before the existing cracks reach the allowable limiting depth. j
- 4. Sucs:ary and Discussion j The analyses presented above have been carried out to establish the technical basis for continued safe operation of the Trojan feedwater system. The allowable length of operation has been determined in terns of the ntanber of hours at hot standby, since that is een the sost severe loadings can occur.
The mode of hot standby operation deterinines the severity of the loadings, dich in turn governs the ntanber of allowable hours at hot standby. The mode of hot standby operation projected for future operation of the Trojan plant will be discussed separately. The allowable hours range fra 800 to 6500 for plants with severe stratification, and an indefinite period is allowable for seat plants, as ahown in Figure 4. Metallurgical erandnations have been conducted in at least 18 sections of feedwoter lines dich were removed fra 12 operating plants during the period 1979 to 1982. The results of these studies have been stamnarized in Appendix A, and show that most of these cracks blunted and became inactive at a depth of approximately 0.100 inches. Only in three or four cases did the cracks propagate extensively, and out of all the PWR plants tich were inspected as a result of IE SJ11etin 79-13 Therefore, fra a practical point of view, there is a very low likelihood that an indication of 0.100 inches detected in a feedwater line after several years of operation would actually be an active or growing crack. This is especially true because the few cracks dich did propagatt significantly did so over a rather short time period, and were l i
-.,_._-..~- ..._ - _- - . - . : -. - _ - - . _ . . - . - _ _ - - - . . . - - _ _ _ . . - . . - . - - . . - - - - - - -
l Trojan Nuclear Plant Document Control Desk Docket 50-344 At t ac hment 2 License NPF-1 April 27 1988
- Page 10 of 24 subjected to such scre severe and more frequent stratification conditions than have probably been experienced at Trojan.
In addition to satisfying the margins.of Section XI for the Trojan feedster lines, there are considerable margins of safety inherent in these lines. Besides the general propensity of the cracks to blunt out at a depth of 0.100 inches as discussed above, there is a deconstrated capability for leak before break in these lines, as evidenced by the Plant B. Examination of the actual flaw profiles in the few cases there significant flaw extension occurred, as shown in Figure 2 and 3, shows that the crack extension is very uneven, which would always lead to a local penetration of the wall, and leakage, long before the flaw length would boccane critical. Study of Figure 1 reveals that the critical length of a through-wall flaw in the Trojan feodater lines is 23 5 inches for the faulted loading, and over 32 inches for norinal operating conditions. Figure 5 shows that considerable leakage would eccur from a through-wall flaw long before it would reach a critical length. This leak rate calculation was done using the approach of reference 7, and the crack opening area was done using reference 8. This approach has been extensively validated experiaantally. In stenary, the integrity of the Trojan feedwater lines has been established for continued operation in compliance with the guidelines of the ASME Code Section II. The most important consideration for continued safe operation with the known indications is thermal stratification during hot standby operation. Based on the expected behavior of the Trojan feedwater syste, at least 6500 hours of hot standby operation is allowable before the presently existing indications would be projected to reach the allowable flaw depth limit. 1 6
- - - - - - -- . _ . - _ . _ _ _ _ , f;r n y - gr:: , _ _, _,_ __ ;_ , ,_ _
, Trojan Nuclear Plant Document Control Desk Docket 50-344 Attachment 2
, License NPF-1 April 27, 1988 Page 11 of 24 5 References
- 1) ASME Section II Task Group on Pipe Flaw Evaluation, Meeting Minutes, 1
January 19,1987
- 2) Zahoor, A., et.al. "Evaluation of Flaws in Ferritic Piping", EPRI Final Report, Project RP 1757-51, October 1985.
- 3) "Evaluation of Flaws in Austenitic Steel Piping" by Section XI Task l Group on Pipe Flaw Evaluation, Trans. ASME Journal of Pressure Vessel Technology, August 1986.
- 4) Bamford, W. H. and Davidson, J.A., "Integrity Assessment of Feedwater Line Indications, Trojan Nuclear Plant" Westinghouse Electric Corporation WCAP 9613, September 1979
- 5) Bamford, W. H. , Thurman, A and Mahlab, M. , "Fatigue Crack Growth in l Pressurized Water Reactor Feedwater Lines", ASME Pressure Vessels and l 1
Piping Conference, Denver, Co. , June 1981, Paper No. 81-PVP-2. I
- 6) ASME Boiler and Pressure vessel Code,1986 edition l
l
- 7) Fauske H. "Critical Two-phase Steam Water Flows", Proc. Ihat Transfer and Fluid Mechanics Institute, pp 79-89, 1961. ;
- 8) Tada, H. and Paris, P.C., "Application of Fracture Proof Design Methods lising Tearing Instability Theory to Nucicar Piping Postulated Circumferential Through Wall Cracks", lENRC NUREG 3464,1983
_----.._,--,,?- - -, .~,, , _ , , . . , - ., __, ,_ _ . , , - -
-.+.n... , . ,.. _, . - , . .
.- Trojan Nuclear Plant Document Control Desk
,- 0-: 1 . 50 2,, Attachment I License NPF-1 April 27, 198
..* Page 12 of 24 Taste 1. ,
},-6310 -l TABLE ALLOWARLE DID-OF-IVALUATION FERIOD FIAW DEFIRI TO TEICDIESS RATIO FOR CIRCU GERINTIAL FLAWS WORMAL OFIRATING (INCLUDING UPSET AND TEST) CONDITIONS f (LIMIT 14AD ANALYSIS) ,
Batio of Flav Imagth, E f to Pipe circumference (Note (3)) (P + Pb)/3 [ Note (2)] m
. .- 0.0 . .0.05 .0.1 0.2 0.3 fa.4 0.5 or
]
. Graater i 1
11.3 0 <75 (4) (4) (4) (4) (4) (4)
~
1.2 0.75 0.21 0.11 (4) (4) (4) (4) 1.1 0.75 0.75 0.53 0.27 0.19 0.16 0.13 1.0 0.75 0.75 0.75 0.44 0.34 0.28 0.23 0.9 0.75 0.75 0.75 0.69 0.48 0.39 0.31 0.8 0.75 0.75 0.75 0.75 0.62 0.50 0.39 0.7 0.75 0.75 0.75 0.75 0.75 0.60 0.46 0.6 0.75 0.75 0.75 0.75 0.75 0.70 0.53 0.5 . 0.75 0.75 0.75 0.75 0.75 0.75 0.60 - 0.4 0.75 . 0.75 0.75 0.75 0.75 0.75 0.66 0.3 0.75 0.75 0.75 0.75 0.75 0.75 0.72 50.2 0.75 0.75 0.75 0.75 0.75 0.75 0.75 I EffES: 1 (1) Flaw depth =.ag for a surface flaw . -- -- -- - -- .- - 2an for a subsurface flaw t = nominal thickness Linear interpolation is permissible - (2) P. = primary longitudinal ammbrane stress (P 10.5 S.) Pb
- Primary banding stress 8 ,= allowable design stress intaasity (La accordance with Section III)
('3) Circumference based se nominal pipe diameter. (4)IVB.3514./shallbeused. Z
- 44
- : - _ -._: - == = = - - _ . . _ - - - . _ - _ - _ -
. Trojan Nuclear Plant Document Control Desk Docket 50-344 Attachment 2
- 1
- uense ri r r - A April 27 1988 l Page 13 of 24 i
'IABLE. 2.
1 - 4sio -T t TARLE i
/ ALIAWABLE END-OF-EVALUATION FERIOD FL4W DEFTRI TO TEICINESS RATIO l f FOR CIRCUHFERDrrIAL FLAWS EMERGENCY AND FAULTED CONDITIONS ]
, (13XIT 14AD ANALYSIS) '
Ratio of Flaw Imagth. Eg. to Pipe Circumference (Note (3)] (P, + Pb 8m (Note (2 0.0 0.05 0.1 0.2 0.3 0.4 0.5 or Greater 12.6 0.75, (4) (4) (4) (4) (4) (4) 2.4 0.75 0.22 0.11 (4) (4) (4) (4) 2.2 0.75 0.75 0.47 0.24 0.16 0.13 (4) 2.0 0.75 0.75 0.75 0.42 0.29 0.23 0.18 1.8 0.75 0.75 0.75 0.59 0.41 0.33 0.25 l 8 1.6 0.75 0.75 0.75 0.75 0.53 0.42 0.31 1.4 0.75 0.75 0.75 0.75 0.65 0.51 0.38 1.2 0.75 0.75 0.75 0.75 0.75 0.60 0.14 1.0 0.75 0.75 0.75 0.75 0.75 0.69 0.50 50.8 0.75 0.75 0.75 0.75 0.75 0.75 0.56 l i NOTES:
- 1) Flaw depth = a for a surface flaw l 2a for a subsurface flaw t = nominal th$ckness Linear interpolation la permissible l
' (2) F, = primary lentitudinal membrane stress (P i 1.0 8.) 4 Pb
- Primary bending stress 5 ,= allowsble design stress intensity (in accordance with Section III)
(3) Circumference based on maniami pipe dismatar. (4) IVB-3514. shall be used. I i 45' wm-g was-w+v-w---v=-t--M w --9 $ w y-wwwem- wy----='N---+-'w-wwwm -g w,-mpwmwy---my---w -e+-- yy-vgwwwe e*= w W"M
,- Trojan Nuclear Plant ,
' Document Control Desk
. Docket 50-344 Attachment 2 License NPF-1 April 27, 1988 i Page 14 of 24 1
\
TABLE 3 RESULTS OF FAT!GUE CRACK GROWTH ANALYSIS i
. l l
Initial Crack Crack Depth after year (inches) Depth (inches) 10 20 30 40 EM E OS 999999 gggggg 0.100 0.1006 0.101 0.102 0.103 0.150 0.153 0.158 0.162 0.168 oo
.
- Trojan Nuclear Plant Document Control Desk j Docket 50-344 Attachment 2 License NPF-1 April 27, 1988
]
j Page 15 of 24 j l
. TABLI 4 1 l
PLAlfT CCMPARISON (REPORTED DATA) l i CYCZ2S NOT SIAEBY H)T I CYCIIS HOURS T AT 107 REALTOR HR/ YEAR STANDBY l YEARS IN OF )CT W OP . HR/EPRY ! OPERATION IFPY STANDBY STANDBY TRIP PLA)ff ,
- : I 2860 99 238 461 Trojan 12 6.2 31
' l 684 145 171 684 Plant F 4 1 171 1
9515 - 9515 13593 Plant B 1 07 55 5, to 6. - 4000 79 571 600 to 800 Plant 0 7 Plant A 9 6.1 1 35 5038 34 560 826 f 1283 - 160 221 r'~ Plant C 8 5.8 132 8203 - M34 9114 Plant D 3 09 30 4 144 1263 100 100 316 l Plant H 7 ) f i i i l l l i i I l i ! J j l
- _ . _ _ _ . . - -__.-- l
, . . Trojan Nuclear Plant Document Control Desk j Dochet 50-344 At t ac hment 2
- License NPF-1 April 27, 1988 Page 16 of 24 l TABLE 5 PLANT C&.PARISON (ANALYSIS BASIS FRCN SITE DATA)
MIN. ANALYSIS HOURS AT 107 STANDBY i HOURS AT HDT STANDBY YEARS IN HOURS PER EPFY PER EPFY OPERATION YEAR IN OP. PLANI HDT STANDB! . . . . l Trojan
- 2860 6.2 461 12 238 l l
Plant B 8678 07 12397 1 12397 6.1 1726 9 192 Plant A 10529 5.8 1557 8 195 Plant c 9029 Plant D 375 8 09 4176 3 1392 , i i
' Reported values for cocparison l / !
l l I
1
..* Trojan Nuclear Plant ,
Document Control Desk ; Docket 50-344 Attachment 2 l License NPF-1 April 27 1988 i
- I Page 17 of 24 Crask Depth, s/t
> - - c.0 c.1 c.4 c.s o.s l 1 I - 1 I i
' gni I
~
t a l l
- f For Port Through Cracks 3
a b h a Thermal + Deadweight f
+ ME \
s - p Thermal + Osadweight 4, nrog _ 106 ,,, + 08 E all Cracks D
~
\ l l
Thermal + Deadweight Onh l 1
,,g i I I i 1 e 10 to 30 40 4 Crook LangA, t unshes) l i i l
i Figure 1: Resuitt of Critical Flaw Size Detemination - Trojan Feedwater Lines
, . . Trojan Nuclear Plant Document Control Desk Docket 50-344 Attachment 2 License NPF-1 April 27,_ 1988 *
- Page 18 of 24 e e *
- e o , *
* "' e
- e. e. ,, ,. \
V' n.w i
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- w w as' w 8"E "'E8" ' men pun m m ee p me **
assium .n, aneum som sessun sy gapet sursaaem em.
/
f
=
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m mass missed w t gap g gemas mass ensema eget 3.tagP 3 Figure 4. Results of Fettsve Crett Greuth Fggure a. Results of Fetteve crest Browth Anales 16 . Plant A. Analr61s . Plant 3 e e e o
".";*2" ,
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w 1 l i l l ! l Fievre 8. Results of Fettese Crest Srewth analrats . Plant c. ! Figure 2: Fatigue Crack Growth Analysis Results Compared with Actual Crack i Profiles (5), Plants A. R and C j
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. Trojan Nuclear Plant , Document Control Desk l Docket 50-344 Attachment 2 ! , License NPF-1 April 27 1988 l
, Page 19 of 24 l l 1 I e e e e . e e * *
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w Figure 5: Leak Rate Predictions for Trojan Feedwater Lines W y - .i
_\l Trojan Nuclear Plant Document Control Desk Docket 50-344 , Attachment 2 j ' License NPF-1 April 27, 1988 Page 22 of 24 g ! CBSDVATIONS_0R.THE GtVTH & FmWATER LTRE CRACKS i l Detailed metallegraphic examinations have been unde on feodater line cracks in over eighteen lines involving tw1ve plants: Zion 1 Beaver Valley 1 , Turkey Point 4 Kewaunee i Palisades Surry 1 l 2 Robinsce Salam 1 DC Cook 1 DC Cook 2 Ginna San oncfra 1 The plants had operated from less than two to more than ten years with the ( asjority havi.ng five to nine */ ears of service. The purpose of this paper is to review the results of the mecallographic examinations in order to draw scue conclusions on the nature of the cracks, their growth characteristics, and the effects of these phenomena on the structural integrity of the feedwter lines. All of the data to be discussed wre drawn free the ICAPS listed at the end of this paper. Free an analysis of the data contained in the ICAPS w can aske the following observations:
- 1. The vast majority of the cracks wre small, of the onder of 100 mils or
/ less la depth. Only three plants contained large cracks, of the order of 500 mils or more in depth. These three plants wre DC cod 1, Seaver Valley 1, and H.B. Robinson. We note that the cracks wre either quite large or very ==11 with practically no internadiate sizes.
- 2. The small cracks wre invariably blunt and filled with oxide and other corrosion products. They tended to have wide a:cths. On the other hand, the large cracks wre relatively clean and had sharp tips. They had grown in a manner independent of the microstructure, i.e. across grain boundaries and second pnases.
- 3. Were they do occur, the large cracks are isolated, with only one or two being seen in any one line. The small cracks are auch more numercus and wide spread, at times covering over 805 of the ciremference of a line.
l J
- 4. Most of the Isrge cracks, including two that were through-unil, wre i
found in a plant (DC Cod 2) that had operated only about one year. l We interpret these observations to mean that m>at cracks initiate early in a j plant's life and then become dornant. The fact that all the small cracks wre 4 shallow, wide, and filled with corrosion products supports this interpretation. It follows that any small crack detected has probably existed for a lors time 4 and is not actively growing. This is further supported by the observation that the large cracks are relatively oxide free and sharp indicating that they grew 1 1 __ __ . _ _ _ _ _ _ . _ _ . . . . - _ _ _ . _ . . _ _ _ _ . _ _ _ _ _ . _ _ _ , _ _ _ _ _ . _ , _ __ _____ _,_.._. m . - -,.. -
. - Trojan Nuclear Plant Document Control Desk
*, , Docket 50-344 Attachment 2 License NPF-1 April 27, 1988
. Page 23 of 24
, . very rapidly. In fact, they grew to a through wall size in less than a year at l
one plant. The observation that the large cracks grow independently of the microstructure suggests that the operating mechanism is low cycle fatigue. Finally, the observation that no intermediate size cracks were found lends credence to the hypethesis that one would have to be extremely fortunate to find an actively gresing crack in its small stage - these cracks simply grow . too rapidly to be detected early. The small detectable cracks are stable and have been so for a ),ong t!ae. On the basis of the preceding discussion, we conclude that the presence of small cracks does not pose a threat to the structural integrity of the feedwater line. The cracks are most likely atable and have existed for a long time. In the event that a crack would initiate and grow rapidly to a large size, experience has shown that it would be local in nature and at worst result in a leak rather than a catastrophic failure. 6 1 l O 1 4 i i I i
. .. . - . - _. = - -
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l ,a i l Trojan Nuclear Plant Document Control Desk
- +
- Docket 50-344 Attachment 2
'* License NPF-1 April 27, 1988 I Page 24 of 24 l- ras. 1 10184 9/82 Zien 1 l 10158 8/82 Beaver Valley 1 i 9385 9/81 Turkey Point 4 9673 .10ng Kewaunee 9669 2/80 Palisades 9638 12n9 surry 1 9616 9/79 HB Robinacn 9591 8/79 Salem 1 9581 8/79 DC code 1&2 9563 8/79 Ginna 9544 San onorre 1 l l d l t l l l ll l l l I I l l l j}}