Text

Letter Attached Thorough Review & Discussion Re Inspection Results of Unit 2 Steam Generator Tubing
	ML19098B418
Person / Time
Site:	Surry
Issue date:	03/21/1977
From:	Stallings C; Virginia Electric & Power Co (VEPCO)
To:	Reid R, Rusche B; Office of Nuclear Reactor Regulation
References
	Download: ML19098B418 (58)
	v • d • e

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e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 2 3 2 61 Mr. Benard C. Rusche Director of Nuclear Reactor Regulation U.S. Nuclear Regualtory Commission Washington, D. C.

20555 Attn:

Mr. Robert W. Reid, Chief Operating Reactors Branch 4

Dear Mr. Rusche:

Serial No.

353B/113076 PO&M/ALH:das Docket Nos. 50-280 50-281 L~cense Nos. DPR-32 DPR-37 The condition, both current and anticipated, of the Surry 2 steam generator tubing has been thoroughly reviewed and is discussed in the attachments to this letter.

The discussions, laboratory tests and examinations, analytical studies, and field inspections have addressed the situation from many different aspects.

A pl_ugging criteria and pattern have been developed and implemented as one result of that effort.

The sum total of the many aspects shows that continued operation of Surry Unit No. 2 can be accomplished within the guidelines of good safety stanq.ards, Because of the multi-area approach of this study, it is believed that a brief discussion of the major points of the attached detailed discussions is necessary to summarize the actions taken and the basis for the conclusion that Surry Unit No. 2 can be safely operated.

The integrity of the tubes has been evaluated for normal operation, for accident conditions, for tubes with leaking defects, for tubes with non-leaking defects, and for defects at tube support plate (TSP) locations.

Tube strength with denting has been shown to be more than adequate for the condition of normal operation and of postulated accidents.

The results consistently show that the operating and emergency limits can be satisfied.,

Extensive examination programs have been conducted.

Eddy-current examina-tions, tube I.D._ gaging, visual inspection, and tube removal have all been comple-ted.

This added information shows good correlation with past, known history.

There have been no findings which introduce a new or unknown situation.

The discovery of three "weeper" leaks (one drop per one to five minutes) which occur-late in the regular shutdown and inspection period was carefully pursued.

The occurrence of this type defect does not in itself constitute a safety concern, but it could be of concern if it is indicative of something of greater potential severity.

Therefore this occurrence was reviewed to determine the cause of fail-ure.

The results obtained from the removal of one of these tubes when compared with the results of three other tubes previou~ly removed has provided assurance that the leaks were caused by the same conditions previously observed and report-ed and not a result of a new or unusual development.

Identification of the "weep-ers at that time*was advantageous in that the defect may have resulted in a lost of

e

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VIRGINIA ELECTRIC AND POWER COMPANY TO Benard C. Rusche P_age 2 plant availability had it_ grown during operation to an extent that the leak limit would have been approached.

Laboratory inspection results have shown that tube defects are an axially oriented intergranular corrosion crack at a TSP intersection dented region.

This typifies a continuation of a known process and does not represent a new or unreviewed phenomenon.

The. defect implications are defined and bounded by our mechanical tube test studies.

Research is proceeding in numerous technical areas.

These programs and their associated content,° progress and schedules* have been discussed periodi-cally with the staff.

The responses being prepared for Surry 1 (Surry 1 SER Feb. 8, 1977) will discuss the overall programs whereas within this letter we have discussed the analytical tube plate strain model in particular detail.

This documents the methods used to anticipate the further extent of denting for the next operating period and the preventative tube plugging program associated with it.

The consequences of tube leakage upon LOCA and upon SLB analysis results have been quantified.

This shows values of a very low percentage of the 10CFR100 limits for SLB, and a very small effect (1°F) on peak clad temperatures for LOCA.

The results also show that they are well bounded by the Surry #1 SER assumptions and calculations.

The details of these accident analyses will be provided with the information requested in Appendix A of the Surry Unit.No. 1 order dated February 8, 1977.

The results of these accident analyses are well within the boundary cases of the staff SER as mentioned earlier.

The probability of occurrence of such a major event coupled with other assumptions taken in the analysis have also been discussed and are shown to be less than 10-7 per year.

The probability would be even lower for the shorter operating time period considerd for Surry Unit No. 2.

The various aspects of continued operation of Surry Unit No. 2 have been evaluated.

The results of the many different considerations consistently sup-port continued operation.

The potential effects upon the tubes have been quantified and are bounded by known facts from test and analysis.

We therefore request that Surry Unit No. 2 be permitted to operate as follows:

1. Unit No. 2 shall be brought to the cold shutdown condition in order to perform an inspection of the steam generators within four equivalent months of operation.

Nuclear Regulatory Commission approval shall be obtained before resuming power operation following this inspection.

For purpose of this requirement, equivalent operation is defined as operation with a primary coolant temperature greater than 350 degrees

- F.

e*

VIRGINIA ELECTRIC AND POWER COMPANY TO Benard C. Rusche Page 3

2. Primary to secondary leakage through the steam generator tubes shall be limited to 0.3 gpm per steam generator *. With any steam generator tube leakage greater than this *limit the reactor shall be brought to the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The leaking tube (s) shall be evaluated and plugged prior to resuming power operation.

3. Reactor operation will be terminated if the primary to secondary leak-age limit of 0.3 gpm per steam generator is exceeded twice in any 20 day period.

Nuclear Regulatory Commission approval will be obtained before resuming reactor operation.

4. The concentration of radioiodine in the primary coolant shall be limit-ed 1 µCi/gram dose equivalent I-131 during normal operation and to 10 µCi/gram dose equivalent I-131 during power transients.

We believe the operation of Surry Unit No. 2 as described above will in no way endanger the health and safety of the general public.

Attachment cc:

Mr. Norman C. Moseley Very truly yours, C. M. Stall=i:ngs Vice President-Power Supply and Production Operations

SURRY UNIT NO. 2 INSPECTION RESULTS A.

EDDY CURRENT TESTING During the February-March 1977 outage, a comprehensive eddy current inspec-tion program was performed in each of the three steam.generators.

This inspection was primarily to determine the sizes of dents in the tubes at each of the tube support plate intersections and to compa+e this with data obtained previously.

It also served to provide the very informative ID gaging data discussed below.

Special attention was given to probing the leaking tubes in order to pin-point the elevation of leakage.

These data were not conclusive, because of the interference given by the dented portion of the tube at the inter-sections.

Improved techniques for identifying possible flaws in dented regi9ns are expected to be available for field trials in late April 1977.

The eddy current inspection program for each steam generator did provide a large quantity of data from which estimates of dent sizes can be obtain-ed.

The inspection pattern consisted of probing each tube through the upper support plate as shown by the boundary line on the included SG Tube*

Maps.

This area coverage encompasses approximately 1000 tubes per steam generator, exclusive of tubes previously plugged.

In addition, two peri-pheral wedge regions were probed in steam generator ZA to discern if any significant changes may have occurred in those regions.

The wedge tubes which were probed are bounded by R28-39, Cl0-23 and R39-45, C22-40.

B.

TUBE GAGING The ed,dy current inspection progrcpn des"cribed above was also performed

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in such a manner as to provide comprehensive data on the tube's ID 1-1

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dimensions.

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obtained using a series of p.roli sizes, viz.,

0.540, 0.610 and 0.650 inch diameter.

The tubes were successively probed with each diameter and recordings made of those which would pass a given probe diameter at which elevation.

The results of this part of the inspection program are given in the enclosed figures 1-1 through 1-9.

The composite results of the three probe sizes shows a definitive pattern of tube constriction. in the area adjacent to the tube lane in each steam generator.

Generally, it is noted that the tube ID increases progressive-ly in moving away from the tube lane.

As will be discussed elsewhere in this submittal, the findings are in agreement with analytical predictions of the finite element analysis mode.

The correlation of this information is used as a basis for preventive plugging, also described in this sub-mittal.

C.

VISUAL INSPECTIONS The tube support plate flow slots of the lowest and the uppermost locations in the Surry 2A steam generator were inspected and measurements taken of the slot openings.

These data are to be compared with similar measurements made during the November 1976 outage.

For the first (lowest) tube support plate, flow slot openings were measured to be as follows:

11/76 2/76 Slot 1 J."

7/8" 2

1/2-3/4 11 7/8" 1-2 3

4 3{8-1/2" --pr L 1/8" 5

I 172-1 3/4" 6

(Manway) 1 7/8 11 Side 1 1/4" 1 5/8"

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e During the iterim period between 11/76 and 2/77, the plant was operated for approximately 1.5 EFPM.

The changes in slot opening are only slight and near the error of measurement.

A similar set of measurements were mad~ of the seventh (uppermost) tube support plate; access being gained by removal of the closure on the 3-inch upper S/G shell penetration installed during the 11/76 outage.

The measurements of slot opening were as follows:

11/76 Slot 2/77 1

1 1/2 11 1 1/2" 2

1 3/8" 1 3/ 8" 3

1 1/2" 1 1/2" 4

5 1 1/2" 1 3/8" (Manway 6

Side) 1 1/2" As expected for the upper tube support plate, not discernable change in flow slot dimension is found.

It was discovered that several locations in the lower TSP of steam genera-tors 2A and 2C along the hot leg flow slot edge were cracked.

These cracks were parallel to the tube lane and between row 1 and row 2 tubes.

This condition was reported to the staff and discussed by phone on February 16, 17, 1977.

This "islanding" phenomenon has been reviewed with respect to accident implications and is discussed elsewhere in this submittal.

D.

INSPECTION OF ALL REMOVED TUBES Four tubes, which reportedly exhibited leakage at dented regions, were re-moved from Surry 1 and 2, and examined in the laboratories at the Westing-house R

& D Center.

The major findings are as follows:

Plant Tube

1.

SURRY 1, S/G C R2C42, HL 1-3 Findings TWO AXIAL CRACKS, 3/4" LONG, LOCATED WITHIN FIRST T.S.P.

INTERSECTION.

INTERGRANULAR AND INITIATING FROM THE I.D.

e Plant

2.

SURRY 1, S/G A

3.

SURRY 2, S/G B

4.

SURRY 2, S/G A Tube R3C63, CL R4C30, HL RSC40, HL

.inding_s NO DEFECTS AT EITHER FIRST OR SECOND T.S.P. INTERSECTIONS NO DEFECTS AT FIRST, SECOND, THIRD, OR FOURTH T.S.P.

INTERSECTIONS.

METALLOGRAPHIC EXAMINATION CONTINUING TO CONFIRM.

AXIAL CRACKS AT FIRST, SECOND AND THIRD T.S.P INTERSECTION PER X-RAY; NO DEFECTS AT FOURTH FIFTH, OR SIXTH LOCATIONS.

METALLOGRAPHY IN PROGRESS; TWO CRACKS AT THIRD T.S.P. ARE INTER-GRANULAR, INITIATING AT I.D.,

SIMILAR TO ITEM (1).

The details of the examination of each tube follow.

EXAMINATION OF TUBE R2C42 REMOVED FROM SURRY UNIT 1, S/G "C" :

The laboratory examination of the section of tube R2-C42 is essentially com-pleted.

Visual examination revealed two longitudinal cracks, each about 3/4" long, centered within the first tube support plate intersection, and separated radially by about 110°.

The two cracks were identified as 160° and 270° respectively, based on an arbitrarily chosen origin and are shown in Figure 1-10 and 1-11.

The examination proceeded as follows:

1.

Double wall Radiography:

The two major cracks showed branching at their centers and at their ends; some additional minor indications were present but were not interpretable.

2.

Diametral Measurements:

A six inch long section of the tube containing the cracked region was positioned in a lathe and rotated under a dial indicator.

Readings were taken at 45° intervals at a number of axial locations centered on the cracked region; these measurements were then plotted to define the tube OD geometry at these various locations.

There is considerable uncertainty in these results because of the deformed condition of the tubes (believed to be caused during extraction) and 1-4

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e the inherent difficulty of establishing a true center line, Despite this, the measurements indicate some flattening and ovalization with some reduction in diameter from the original 0.875".

3.

Metallographic Evaluation:

A two inch long section containing the cracks was removed and. split longitudinally into two halves, each con-taining one of the cracks.

0 Metallography was performed on the 270 crack by sectioning through five transverse locations and examining be-fore and after etching.

a.

Crack Morphology:

The crack was intergranular in nature and ini-tiated from the ID surface.

b, Structure:

The material was very fine grained, ASTM 9-10 grain size, with noticeable amount of intragranular precipitates, both at the cracked region and at an unattacked location 16" away.

c.

Hardness:

The hardness at the cracked region was RB 95 to RC 23 as compared to RB 88 to 94 at the unattacked location.

d, Wall Thickness:

Measurements taken at several locations around the periphery in the region of the crack indicated thickness within the normal tolerance band.

e, Scanning Electron Microscopy (SEM):

This examination also revealed the intergranular nature of the crack and the numerous intragranular precipitates previously observed in the optical microscopy.

4.

Chemical Analysis:

A bulk analysis of the tube material indicated a normal composition:

C:

0.046%

Fe:

7.1%

Cr:

15.2%

Ni:

74.8%

1-5

A preliminary ED~ analysis from the fractured s~ace of the 270° crack was not definitive.

Somewhat higher than background readings were ob-served for certain rare earth elements and Iodine, judged to be fission product residue.

0 Fractography has been started on the 160 crack.

The initial results confirm that this crack is also intergranular in nature and originating from the ID.

SEM examination of the fracture face indicated a relative-

... ly clean surface with no definitive aspects.

EDAX analysis of several adherent particles indicated typical Fe, Ni and Cr.

The examination of this crack is continuing, incl;uding a planned Spark Source Mass Spectro-metric analysis of the fractured face.

The results indicate a fracture mechanism similar to that of the U-bend cracks, that is, a stress induced intergranular attack, but a quantitative description, including the role of strain and strain rate,' awaLts further tests and analysis.

EXAMINATION OF TUBE R3C63 REMOVED FROM UNIT 1, S/G 11A 11 A 13' length of Tube R3C63 was removed from the cold leg side and sub-mitted for laboratory examination.

The removed portion, which was in six sections, contained the first and second tube support plate inter-sections; leakage reportedly had occurred at the second tube support plate location.

A cut had inadvertently been made through this location at extraction; however, other than precluding a dimensional and eddy current examination of that dent, it did not prevent a meaningful exami-

nation, Visual examination did not disclose any apparent defects in the two sections containing the tube support plate intersections, nor did eddy current examination of the first support location or radiography of both locations.

1-6

e Dimensional measurements were taken below, through, and above the first support location:

considerable ovality was detected in and above the support location.

These measurements are currently being analyzed and are to be related to the strain intensity plots of the support plate.

Metallography was performed by sectioning through the remaining region of the second support plate location.

No cracks or other detects were observed on either the ID or OD surfaces.

The structure was clean, with a fine grain size similar to that observed in R2C42.

Hardne.ss traverses are planned but have not been performed to date, It is concluded that the reported leak must have occurred in a location above the second tube support intersection.

EXAMINATION OF TUBE R4C30 REMOVED.FROM SURRY UNIT*2*s/G "B" A number of sections of tube R4C30, containing the first four tube sup-port plate intersections were submitted for examination.

The visual examination did not reveal any obvious defects or abnormal conditions.

Both eddy current and radiographic examinations of the four dented regions were negative, as were the metallographic sections taken through the center of each dent.

The microstructure indicated a somewhat coarser.*

but normal grain size (ASTM 6-7) than previously observed in the first two tubes examined.

Hardness traverses are to be taken and related to the measured deformation.

Detailed diametral dimensions were taken through each dent and these measurements are currently being analyzed.

This information is expected to be helpful in defining the critical deformation associated with the initiation and propagation of cracks.

As in the case of Tube R3C63, it is concluded that the reported leakage must have occurred in the tubing above that portion which was removed for examination.

1-7

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EXAMINATION OF ~y 2 TUBE RS C40 HL S/G A A length of tube RS C40 from S/G A of Surry 2, which reportedly exhibited leakage, was removed and submitted for examination.

The length which was cut during extraction into 11 sections, each 2 ~o 4 feet long, comprised the portion of the hot leg from the primary side of the tube sheet to above the sixth tube support elevation.

The laboratory examination consists of visual, dimensional, eddy current,

-double wall radiography, and metallography.

The initial results are as follows:

1-8

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't VISUAL EXAMINATION OF TUBE O.D.

The sections appear generally similar to previous tubes which have been extracted; the O.D. surface exhibited. a uniform dark oxide, except at the tube support plate locations where the surfaces are blotchy, (found in other tubes to be caused by phosphate deposits).

There was one open axial crack found at the third tube support plate location, but no other obvious cracks were observed.

There were a number of deep axial gouges and scratch marks along the lengths, and general ovalizatiqn and bow.

Figure 1-12 shows the external appearance of the visible crack.

DIMENSIONAL EXAMINATION The O.D. Dimensions at each dented region were measured by carefully position-

ing each section in a lathe and rotating the tube under a dial indicator.

Continuous traces were obtained of the deviation from nominal radius as a function of angular rotation and these traces were converted to O.D. con-tours.

These traces were taken at l/8 11 axial intervals through the 3/4" long tube support plate intersection and at 1/4" intervals for about one inch above and below the dented region.

In the case of the third tube support location, which contains the visible crack, a replica of the O.D.

was made by casting a flexible molding material over the tube; after curing, the mold was stripped off and a positive casting will be made which accurate-c...

ly reprodu71-the actual tube dimensions and contour.

The measured dimen-sions and the replica will be used to relate the contours to the crack locations and, hopefully, to hoop strains in the tubing.

This work is under-way.

EDDY CURRENT EXAMINATION Each of the six sections* which contained.a tube *support plate intersection was e~amined with a 0.540" dia. probe at a frequency of 400 KH.

A large z

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crack signal was observed at the 3rd T.S.P. location, which obscured the normal dent signal.

No other identifiable crack signals were observed at any other location.

DOUBLE WALL RADIOGRAPHY Each dented region was X-rayed at four angular orientations with a technique capable of 2% wall thickness sensitivity.

Axial crack indications were de-tected at the first, second, and third tube support plate locations; no identifiable defects were observed at the fourth, fifth, or sixth locations.

Radiographs of the third tube support plate location indicated two axial cracks; those of the second location revealed several short (less than 1/4" long) axial cracks.

A single axial crack, perhaps 1/2" long, was observed at the first intersection.

METALLOGRAPHY A transverse section was cut through the cracked portion of the third support plat~ location to determine the mode of failure.

The section was taken approximately 1/4" from the bottom edge of the support plate intersection.

Figure 1-13 shows the section after rough grinding.

The O.D. is deformed concavely at the location of the major crack.

Polished and etched micro-graphs of the major and branching cracks (Figures 1-14 and 1-15) indicate that the fractures initiate from the I.D. surface and propagate in an intergranular mode, similar to the previously examined R2 C42 Surry 1 tube and the various Row 1 U-bends from Surry l and 2 and Turkey Point 4*, -Thus the same mechanism of failure, that of strain enhanced stress corrosion cracking, is evidenced.

Additional metallography is currently underway with sections being taken through *each tube support plate location.

These are expected to provide additional insight into the critical deformation associated with cracking 1-10

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in dented regions, since some locations are presumably uncracked, and others have part-through wall defects.

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'([,,

" TUBE PLUGGING PATTERN AND ANALYSIS The basis of the analysis has been the finite element model which is shown in figure 2-1.

With the aid of this model, plate expansions have been considered up to the full closure state and for a significant period beyond.

The speci-fied closures of flow slots of various tube support plates for all Surry I and II steam generators are shown in Figures 2-2a and 2-2b.

Linear closure rates are assumed based on Effective Full Power Months since AVT chemistry changeover, and are averaged over the six flow slots for any given plate; The maximum clos-ure rates appear on the lowest plate with an average rate 0.15 in. per EFPM (AVT) and a maximum of 0.17.

Based on this available data, these rates are representative of all the Surry I and II plates except the top plate in each which exhibits an average rate of 0.08 in. per EFPM (AVT) and maximum of 0,11 inches closure per EFPM (AVT).

Using the closure rates above, a correlation of EFPM (AVT) and expansion co-efficient for the support plate computer model was obtained.

The expansion.

sequence of full flow slot closure uses the assumed maximum closure rate of

.17 in. per EFPM (AVT) based on the results of Figures 2-2a and 2-2b.

Figure 2-3 illustrates. the sequences of runs to 5 months beyond closure for Surry II.

The strain intensity plots and equivalent strain plots for all runs that were made are shown in Figures 2-4 through 2-11.

Figures 2-4 through 2-9 show the location of tubes that have leaked with respect to strain intensity and equiva-lent strain at approximately the extent of flow slot closure at which the leaks occurred in both Surry I and II.

F~om this data we have arrived at a lower bound of plate strain intensity (and related equivalent strain) of 0.08 in/in (and 0.045 in/in).

To establish this number, the distribution of leaking tubes 2-1

~

with respect to strain intensity (and related equivalent strain) has been examined as illustrated in Figure 2-12.

The data indicates that 95% of all leaking tubes have occurred at locations of plate strain intensittie~ of 0.08 in/in or greater (> 0.045 in/in equivalent strain).

Tests have been performed to correlate these plate strain intensities and equivalent strains to tube hoop strains and equivalent strains.

The results of the first tests are shown in Figure 2-13.

The initial test results indicate the magnitude of hoop strain expected in the range of plate strain intensities and equivalent strains that are of interest.

Table 2-1 shows the tube equiva-lent strains that develop as a result of these hoop strains.

It appears that the equivalent strains in tubes containing cracks at dented regions are some-what lower than in the small radius U-bends.

But is also should be noted that the hoop strains are much higher in the dented region than in the U-bend.

It is also believed that constant straining of the dented region is a facto; in the cracking phenomena. The population of tubes that lie within the lower bound is shown in Figure 2-14.

The large majority of tubes within this "strain boundary" are not expected to encounter leakage.

The tubes that should be plugged are those for which the probe inspections in-dicate severe tube deformation.

Most investigations have indicated that stress corrosion cracks at "leaky dents" occur in tubes with a minmum inside diameter in the range of 0.500 inches.

Locations of previous leakers and those locations that cannot pass.540 and.610 inch probes will give extensive information on the progression of the area of deformed tubes.

Based on the strain intensity and equivalent strain plots, the growth of the contours in the range of.08 in/in to.12 in/in (the range in which over 90% of the leakers have occured) is about 1/2 tube row per month in the region of the strain boundary.

The pattern 2-2

of tubes with large deformations follow these contours.

A population of de-formed tubes will be selected in order to determine the length of time for which a given plugging pattern removes from service those tubes with the highest poten-tial for leakage.

This length of time is related to the 1/2 tube row per mont_h :3rowth of the strain contour.

For purposes of conservatism, it is assumed that a tube which does not pass a O. 540 inch probe is vulnerable.to leakage (limited data show that leakage has occurred in tubes with a minimum diameter considerably below 0.540, i.e,,

~0.470 inches).

Based on the 1/2 tube row per month, each additional row in the affected region that is plugged affords two months of operation.

Inspection data indicates that this region is populated with tubes which will also not pass the 0.610 inch probe.

Thus, for additional conservatism, the tubes in the affect-ed region that do not pass the 0.610 inch probe will also be plugged.

The information discussed above provides the basis to formulate a plugging pattern for the Surry Unit 2 steam generators.

With the implementation of t?is plugging pattern, the unit is expected to operate trouble free for a period in excess of four months.

The criteria is:

a)

All tubes which do not pass the 0.540 inch probe will be ulugged.

This results in plugging approximately 15 in 2A, 14 in 2B, and 14 in 2C.

b)

Additionally, for four months operation, two tubes beyond (i.e., high-er row numbers) all tubes in any column found not to pass the 0.540 inch' probe will be plugged.

2-3

c)

All tubes which do not pass the 0.610 inch probe will be plugged.

This results in plugging approximately 61 in 2A, 30 in 2B, and 25 in 2C.

d)

All tubes plugged in a) and b) in each column will be plugged to the tube lane, if not already plugged.

e)

As a conservative measure, tubes completely surrounding (including the diagonal) all known leaking tubes will be plugged, if not covered by a), b) and c).

As a result of this program, the tube plugging levels will be as shown on figures 2-15 through 2-17.

This equates to 16.1% for 2A, 16.6% for 2B, and 16.4% for 2C and 16.4% for the unit.

2-4

TABLE t-1 TENTATIVE TUBE STRAIN VALUES

-. 12 ~

2 e: = -

(e:

e: ) +

3

.H A

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= ~-

=* tube axial strain A

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(

2

) = radial strain Plate Strain Intensity(%)

e:

1 e:2 8(1) 10 12 Plate Equivalent Strain (%)

4. 5 (l) 5.6 6.7 1/2 Tube Hoop

Strain(%)

e:H

.7.o 8.8 10.0 Tube Equivalent

Strain (%)

-e:

7.3 9.2 10.4 (l)Based on strain intensity and equivalent strain plots, these are representative relative values of the two quantities.

2-5

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60.

45.

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Monel Characteristics

1.

Hot/Cold Side Expansion Bias

2. Elastic Behavior of Channels at Support Locations (Channel stiffness was obtained from crush tests performed at Westingho 1.1se)

3.

Wrapper Stiffness Incorporated at Periphery

4.

Effect of Cracked Ligaments - Tube Stiffness

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2-14 L

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2-18

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Relationship Between Tube Hoop Strain and Plate Strains

I.

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0 if,)C;-ilculated strain at crack

'~ in* Tube RS C40 as related to strain intensity and equi.valent strain maps, of Figures 5-8 and 5-9 e>

6.

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COLUMNS 94 9?. 90 88 Bio M 82 80 78 11,, 74 72 70 ~8 l>" /.,A ~2 t,,IJ 58 5t,, 54 52 5o 48 41. 411 42 '10 31:l 31,, 3'1 32 30 ZB 2~ zq 22 20 /8 ft,, /4 /Z /0 B l.i, 4

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ATTACHMENT 3 NRC COMMENT "Provide fluid/structural analysis of inner row tubes, considering loss of one, two, and three lateral supports due to the so-called "islanding" phenononon caused by support plate fracture parallel to the flow slots."
. RESPONSE Under operating conditions, recirculated water and feedwater mixtures enter the tube bundle through a 14 inch high opening between the tubesheet and the wrapper barrel.
The cross-flow velocity U in the gap between the outermost tubes is given by:
u C W P 8.7 ft/sec.
y (2nRH) (P-D) where:
C circulation ratio, 4.5 W
steamflow rate, 3.9(10)6 lb/hr y
fluid density, 48 lb/ft3 R
radius of outermost tube, 5.00 ft H
wrapper opening height, 1.17 ft.
P tube pitch, 1.281 inches D
tube diameter,.875 inches For tubes both in the vicinity of the inner row and the wrapper, tube lane blocks could increase the flow velocity locally by approximately 5% above that calculated above based on three dimensional flow calculations.
Hence, the maximum cross-flow velocity for the outermost tubes, closest to both the wrapper and tube lane is 9.2 ft/sec.
The corresponding cross-flow velocity is less for tubes in the interior of the bundle.
3-1
\\ 'Flow-induced vibration analyses of inner row tubes were performed to evaluate the effects of various support conditions.. The mathematical model considered is shown in Figure 3.-1.
The tube is fixed at the secondary face of the tubesheet, and the unsupported span length (1) was incremented in steps of 50.5 inches to simulate one, two and three missing support points.
Hence, results are provided for L = 50.5, 101.0; 151.5 and 202.0 inches.
The case, L = 50.5 inches, represents the nominal or full support condition.
Both vortex shedding and fluidelastic tube vibration mechanisms were considered for 100% load operating conditions.
Even with conservative assumptions, a fluidelastic vibration mechanism cannot be established under conditions of full support or with one missing support.
With the additional removal of two or more lateral supports, a fluidelastic vibration mechanism is still not expected.
Variations in the support conditions for adjacent tubes in the vicinity of a given flow slot lead to correspond-ing variations in natural frequencies from tube to tube. Test results have shown that frequency mismatching of the tubes causes a significant increase in the critical velocity required to set up a fluidelastic vibration.
Therefore, detailed calculations were carried out for the vortex shedding vibration mechanism only.
When fluid flows in cross-flow past an array of tubes, vortices appear alternately on one side and then on.the other side of each tube.
Upon reaching a maximum size, the vortices detach and move downstream in the tube wake.
For stationary tubes, the shedding of vortices occurs at discrete frequencies which depend on the flow speed, the tube diameter, and a dimensionless parameter called the Strouhal number.
Because of the shedding of the vortices, the tubes experience an alternating force in a direction perpendicular to that of the flow.
Since the magnitude of the alter-nating drag force is approximately on tenth that of the alternating lift force, this report is focused on the response of the tubes to the alternating lift forces orily.
3-2
) *
~ In order to calculate the vibration amplitude of the tubes for a given cross flow velocity distribution, it is necessary to know(l) the magnitude of the alternating lift forces, (2) the frequency at which the vortices are shed, and (3) the damping in the tube bundle.
The alternating lift force per unit length is given by:
F C (pU2/2) Drr sin(2rrfst) where:
F = alternating lift force per unit leng.th C = alternating lift force coefficient p
fluid mass density u = cross flow velocity fs vortex shedding frequency, cps t
time, sec.
D tube diameter For a stationary tube subjected to cross flow, the vortex shedding frequency is:
where:
SU
-.D-S Strauhal number When the vortex shedding frequency approaches a natural frequency for a flexible tube, however, the vortex shedding becomes synchronized with the tube's natural frequency.
3-3
In the calculations summarized in Table i-1, the vortex shedding frequencies were assumed to occur at the fundamental frequency for each tube, and damping was assigned the low value, 1%.
Even with these conservative assumptions in effect, the results predict only moderate tube vibration amplitudes and bending stresses with lateral support missing at one or more tube support plates.
Only in the case when three supports are missing, does the calculated tube vibrating amplitude exceed one half of the gap between adjacent tubes.
Hence, excessive tube wear resulting from tube/
plate contact or tube/tube contact is not expected during operation.
While a precise determination of the maximum wear rate cannot be determined, it is expected to be considerably less than the maximum tube wall thinning rates experienced recently in the U-bend region of the San Onofre steam generators.
Fretting of tubes at carbon steel anti-vibration bars in those units was attributed to excessive gaps between the tubes and bars.
Although the cross-flow velocities calculated here for the tubesheet region are approximately equal to those in the U-bend region at San Onofre, they are directed locally over a relatively short length of tube near the secondary face of the tubesheet where the tube is fixed.
A review of historical eddy current test data from inspections at San Onofre indicates maximum wear rates were less than 40% per year.
Based on field evidence and the analytical results presented above, it can be con-cluded that flow forces on representative inner row tubes would lead to negligibly small tube vibration amplitudes and bending stresses.
In addition, excessive tube wear is not expected during operation.
The relationship of the actual situation of the Surry 1&2 tube support plates to the vibrational analysis must be recognized.
The hard spot corner cracking and 3-4
~ subsequent potential for loss of lateral support has not been seen to be extensive and also has been limited to the first tube row.
Little effect extending to row two is anticipated,and none extending to row three. Only the first column (50.5) of Table 2-1 would therefore apply to row three and greater.
Also, it should be noted that the rows in question, row one and row two are no longer in service in Surry Unit #2 as a result of prior plugging.
Additionally, most of the tubes in Row 3 are also plugged where flow-slots are located.
3-5
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Dear Mr. Rusche:

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