ML19093A142

Inservice Inspection Program, Refueling Outage No. 2, Surry Power Station Unit No. 1, Report No. ISI 75-8

ML19093A142
Person / Time
Site:	Surry
Issue date:	12/30/1975
From:	Virginia Electric & Power Co (VEPCO)
To:	Office of Nuclear Reactor Regulation
References
ISI IR 1975008
Download: ML19093A142 (11)

Text

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I I REGULATORY DOCKET FILE COPY I INSERVICE INSPECTION PROGRAM REFUELING OUTAGE NO. 2 I SURRY POWER STATION I UNIT NO. 1 I DECEMBER 30, 1975 I REPORT NO. ISI 75-8 I

DOCKET NO. 50-280 I LICENSE NO. DPR-32 I

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I Vepco

,. VIRGINIA ELECTRIC AND POWER COMPANY I

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I INSERVICE INSPECTION PROGRAM REFUELING OUTAGE NO. *2

.I SURRY POWER STATION I UNIT NO. 1 I. DECEMBER 30, 1975 I

REPORT NO. !SI 75-8 I

I DOCKET NO. 50-280

_LICENSE NO. DPR-32 I

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I Vepco I

I VIRGINIA ELECTRIC AND POWER COMPANY I

I I I. INTRODUCTION In accordance with the requirements of Technical Specification I 6.6.C, this report summarizes the results of the inservice inspection I activities performed during Refueling Outage No. 2 of Unit No. 1 at the Surry Power Station.

I The document entitled Inservice Inspection Program, Refueling Outage No. 2, Unit No. 1, Surry* Power Station, Report No. !SI 75-4 dated July 1975 I provides- the specific details concerning the inspections which were scheduled I to be performed.

The inservice inspections were conducted by Virginia Electric and Power I Company (Vepco) representatives. The areas inspected by Vepco personnel I are detailed below:.

Tech Spec Area Method of I Com~onent Refer. Inspected Inspection Low Head Safety 7.2 Piping in valve VT I Injection System Piping pit I II. INSPECTION

SUMMARY

Inservice inspections-:for Unit No. 1, Refueling Outage No. 2, were con-I ducted in accordance with Inservice Inspection Repqrt IS! 75-4. The low I h~ad* safety injection system piping located in the valve pit was visually inspected during the period November 4-9, 1975.

I In addition to the inspectiqns required by the Technical Specifications, eddy current examinations of the steam generator tubes were conducted. The I results of these inspections are reported herein.

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I III. INSPECTION RESULTS The visual inspection of the low head safety injection system piping I in the valve pit indicated satisfactory results. Many of the previously reported arc* strikes had been removed. Visual indications of minor rusting I and corrosion were noted. Leaky packings and gaskets were repaired to I eliminate sources of leakage *

. Extensive examinations of the steam generators were*conducted to sat-I isfy the requirements of Regulatory Guide 1.83.and to obtain additional data on the generic tube diameter reduction phenomenon,. i.e. denting. In ad-I dition to eddy current testing of the tubes, a visual inspection of the I U-bend region of one steam generator was conducted. Additional visual in-spections of the tupes and tube support plates were conducted through hand I holes on all.three steam generators and on "hillside" ports on A steam generator.

I The eddy current testing showed that all* inspected tubes on both the I hot and cold leg sides exhibit some local diameter reduction at most sup-port plates. The extent of this deformation is mostly in the range of less I .

than 20 mils reduction in diameter, but some signals indicate as much as approximately 50 mils reduction in diameter. Deformation tended to be I greatest.in those rows closest to the divider lane (rows 1 through 3) where I gauging of the leaking tubes indicated an inside diameter reduction to 0.550 inch (at the second support plate). No significant differences be-I tween the hot and cold legs were noted.

Visual inspection of the U-bend region and top of the tube bundle of I C steam generator revealed no unusual conditions near the anti-vibration I bars (AVB's) such as shifting, visible corrosion, or tube damage. There I

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I was a hard deposit noted at the intersection of the AVB's and the tubes.

I There was no apparent shifting of the uppermost (7th) tube support plate, with normal clearance between the support plate and wrapper evident.* The I spaces between the tubes and the tube support plates appeared to be filled with a deposit as no visible gaps were apparent. The tube bundle appeared I to be in its as-built condition with no .visible* bending, bowing, or mis-I alignment of tubes in the U-bend region.

Additional visual inspections through hand holes of A, Band C steam I generators were conducted. In-plane "hour-glassing" or the rectangular flow slots in the support plates in all three steam generators was noted I with the least amount occurring in steam generator B. No signs of abnormal I surface sca~e, deformation, cracks, or out of plane distortion were evident.**

Deposits were evident in the spaces between the tubes and tube support plates I with no clearance between the tube and tube supports indicated. Looking down the divider lane, the tubes appear straight from the top of the tube I sheet to approximately 15 to 20 inches below the first support plate with I gradual bowing from there to follow the _!:ontour of* "hour-glassing" in the support plate flow slots. Uniform spacing was observed between adjacent I columns of tubes.

A summary of the inspection conducted on each steam generator is I given below.

I A STEAM GENERATOR A total of 1583 tubes were checked at 400 KHZ in the inlet side of I which 102 tubes were in the low flow area. These tubes were checked up to and past the 7th tube support on the inlet side and into the U-bend.

I All remaining tubes were checked past the 2nd tube support. Of the total, I

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I I 1477 had indications of tube defects. A breakdown of the indications are as follows: 1185 tubes ~20% defect, 69 tubes 20-29%, 51 tubes 30-39%, 133 I tubes 40-49%, and 39 tubes .::_50%.

I After completing the 400 KHZ tests for defects, 486 additional tubes were inspected in the inlet side and 210 tubes were inspected in the out-I let side. The tes~s were performed *at 400 KHZ with a .700 inch diameter probe and a .540 inch diameter probe to determine the extent of denting in I the steam generator. Of those tubes tested, a .540,inch diameter probe I would not reach the U-bend area in 36 tubes on the inlet side and 31 tubes on the outlet side.

I There are a total of 297 tubes (8.8%) plugged in "A" steam generator.

B STEAM GENERATOR 1*

A total of 609 tubes were inspected in steam generator "B" inlet at I 400 KHZ of which 71 tubes were in the low flow area.

• These tubes were checked up to and past the 7th tube supp.ort on the inlet side, and into the I U-bend. All remaining tubes were checked past the 2nd tube support on the I inlet side. There ,W'ere 76 indications in "B" inlet.

indications are as follows:

A breakdown of the 35 tubes <20% defect, 18 tubes 20-29%, 21 tubes I 30-39%, 2 tubes 40-49%.

Four (4) tubes were explosively plugged in "B" generator. Two (2) of I those tubes were found on this inspection to be 40% and two* (2) additional I tubes were plugged because of arc strike*s on. the secondary side.

An additional 168 tubes were inspected in "B" steam generator for I denting. All tubes inspected would pass a .700 inch diameter probe past the 7th support and into the U-bend area.

I There are a total of 8 tubes (0.2%) plugged in "B" steam generator.

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I I C STEAM GENERATOR A total of 616- tubes were inspected in steam generator "C" inlet I at. 400 KHZ of which 88 tubes were in the low flow area. These 88 tubes were checked past the 7th tube support on the inlet side and into the I U-bend. All the remaining tubes were checked past the 2nd tube support I on the inlet side.

Of the tubes inspected there were 174 -indications of tube defects.

I There were 8 tubes <20% defect, 14 tub~s 20-29%, 43 tubes 30-39%; 54 tubes 40-49% and 55 tubes >50%.

I Because of defects, 115 tubes were explosively plugged in "C" steam I generator.

defective.

109 of those tubes were found on this inspection to be >40%

Six additional tubes were plugged that had indications <40%.

I IV. CONCLUSIONS I The results of the inservice inspectidns performed on the low head I safety injection system piping verified the integrity of the systems and components examined and satisfied the requ-ireni.ents of the Technical Speci-I fications. The discrepancies noted were corrected.

Based on the results of the inservice inspection program, as summarized I herein, the low head safety 1njection system piping inspected has not I experienced degradation and there is reasonable assurance that it will continue to perform*its design function in a safe and satisfactory manner.

I The eddy current inspections of the steam generators indicate that there is significant denting occurring. Inspections of other Westinghouse I steam generators indicate that denting is a* generic concern. The.evaluation I and tentative conclusions as to the nature of this phenomenon are, nec-essarily, preliminary and are based upon currently available information, I

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I including the information presented herein and data obtained from other facilities. Additional data is required to substantiate, or possibly modify, I the postulated mechanism for tube deformation.

The current postulate to account for denting which best satisfies the I physical evidence is as follows: A substance fills the gap between the I tube and tube support plate hole, denting the tube and dilating the hole (mainly through stretching of the ligaments which surround the tube hole).

I The pressure neces,saryto stretch the ligaments is approximately 6,000 to 7,000 psi. As every tube hole dilates, the entire projected area of the I tube support plate increases. At rigid regions of the support plate, the I in-plane displacement of the plate is locally constrained resulting in 1

.'hard spo:ts" which do not contain the array* of interstitial flow holes found I elsewhere in the support plate. The effect of constraint at the "hard spots" is to concentrate the deformation of tubes, tube holes, interstitial flow I holes and flow ports (located at the center of the tube bundle). The de-I formation of the tubes observed at- "hard spots" is more extensive arid of greater magnitude than it is elsewhere in the support plate.

I Immediately after the discovery of local tube deformations in the area of the tube support plates, an extensive analytical and experimental in-I vestigation was started by Westinghouse and supported by the utilities.

I These analyses were based on several mechanisms, all of which were believed to locally reduce the outside diameter of the tube. Of these mechanisms, I the two most probable were thermal ratcheting and crud deposition~ The thermal ratcheting was assumed to occur with an internal to external pres-I sure difference of 1500 psi and an alternating tube OD temperature from I approximately 590 degrees to 490 degrees. This.mechanism alone, however, was shown not to be credible both experimentally and analytically. The I

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I crud deposition model, however, did show credibility in that significant deformation could be obtained in relatively few cycles. The crud deposit I in the tube to tube support hole annulus was assumed to be incompressible and with a zero coefficient of thermal expansion would continue to fill the I gap after each cycle. The total radial deformation of each cycle was I 1.7 x 10-r+ inch. This or a similar mechanism will result in a net pressure on the tube outside diameter and the tube hole inside diameter.

I Determination of*Externai Pressure to Yie"id Tube and Support Plate The tube AP can be approximated with the following relationship:

I AP= (Yield Strength)

+ 1.

,I Mean Radius Wall Thickness 2

I AP= (44,660) = 4620 psi

.4125 + 1

.045 2

'I This pressure difference will cause yielding and a permanent deformation I of the tube. With the internal (reactor coolant) pressure at 2,235 psi, the outer pressure for a AP of 4,620 is 6,855 psi. The 6,855 psi represents the I total "hydrostatic" pressure acting on.the OD surface of. the tube and would include the possible effect of secondary side pressure. When 6,855 psi is I considered to*be applied to the outside of the tube, it is also applied to I the ID surface of the hole in the support plate under the assumptions that the crevice is filled with an incompressible substance.

I The stress across the minimum width membrane in the support plate is calculated to be 35,718 psi, this is greater than the ASME code value at I 500 degrees F for the support plate material of 24,500 psi. Therefore, I the tube support plate will receive a permanent distortion. Actual in-plane distortion of the tube support plates has been observed which tends I to support the latter hypotheses.

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I Mechanical tests are being run to simulate severely deformed tubes and to evaluate the hydrostatic collapse characteristics *of tubes with I given configurations of deformation which are representative of. tube con-ditions observed in steam generators on Surry Unit No. 1.

I The deformation of tubes and support plates observed to date is believed I to be the result of the growth of deposits and/or corrosion product within the support plate in those steam generators which have operated for extended I periods on*phosphate water chemistry and more recently on all v6latile treat-ment (AVT). A substance accumulates in the annular gap between tube and I support plate hole. The substance exerts force directed radially inward on I the tube and outward on the support plate when the gap is filled.

sult is a reduction of tube diameter and an increase of support plate The re-I hole diameter. At rigid* areas of the support plate, the deformation is most extensive and least symmetrical (about. the axis of the tube and of I the .hole).

I Based on all the information on the tube diameter reduction phenomenon available at this time, summaries of which have been presented above, the I following conclusions have been reached:

1. The phenomenon appears to be restricted to units which had I extensive operation with phosphate treatment prior to con-I version to AVT.

2. The phenomenon is hypothesized to be caused by a build up I of a corrosion product in the space between the tube and the tube support plate.

I 3. There is no apparent change in the mechanical properties I of the tube metal. Data from tests performed on tubes removed from steam generators indicate properties (e.g.

I ductility) which are typical of new tube material.

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I 4. Station operation can continue safely while further evaluation and investigation of the phenomenon proceeds.

I 5. Vepco, in conjunction with the Westinghouse Electr.ic Corporation, will continue to investigate the matter.

I Additional eddy current examinations are scheduled during I forthcoming refueling outages.

The reader is referred. to the documents listed in Section V for I more specific details.

I V. REFERENCES*

I 1. Report of Inservice Inspection Conducted by Vepco Personnel, Refueling Outage No. 2, Unit No. 1, Surry Power Station, December *30, 1975, Virginia Electric and Power Company.

• I 2. Inservice Inspection Program, Refueling Outage No. 2, Unit No. l, ISI 75-4, July 1975, Virginia Electric and Power Company.

I Section 4.2, Technical.Specifica1=ions, Surry Power Station, Unit Nos. 1 and 2, Virginia Electric and Power Company.

I 4. Letter to Dr. Robert E. Heineman, Director, Division of Safety Systems, U.S. Nuclear Regulatory Commission, dated December 18, I 1975, NS-CE-871,' from C. Ercheldinger, Westinghouse Electric Corporation (Non-proprietary).

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