ML20235A390

Metallurgical Rept Insp Sequoyah Nuclear Plant Unit 2 Insp of Crossaround Piping,Target Tees & Extraction Steam Piping for Wall Degradation

ML20235A390
Person / Time
Site:	Browns Ferry, Sequoyah, 05000000
Issue date:	11/30/1984
From:	Woods T TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML082420107	List: ML20234F497 ML20234F508 ML20234F519 ML20234F590 ML20235A015 ML20235A154 ML20235A160 ML20235A171 ML20235A203 ML20235A218 ML20235A310 ML20235A347 ML20235A390
References
NUDOCS 8709230291
Download: ML20235A390 (79)

Text

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M ETA LLU RGIC AL R EPORT L C.

INSPECTION SEQUOYAH NUCLEAR PLANT UNIT 2 i

INSPECTION OF CROSSAROUND PIPING, TARGET TEES, AND EXTRACTION STEAM PIPING FOR WALL DEGRADATION 9

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Prepared By

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1 0709230291 070910 ADOCK0500g9 PDR G

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Introduction

- Inspections were performed on the subject piping during the unit 2 cycle 2' f-refueling outage to determine if wall degradation, resul*ing from steam i

erosion, was present.

The inspections consisted of a visual inspection of the crossaround piping

. and target tees and an ultrasonic test (UT) inspection of the extraction steam. piping. Unanticipated failures of this piping could result in injury to. plant personnel and damage to the plant components.

Discussion Visual inspections were performed on the inside surface of the carbon steel crossaround piping to:the 2A2 and 2B2 moisture ser:arator reheaters.

The inspections showed areas of minor erosion on the pipe wall downstream of the turning vanes on each pipe; however,~no appreciable na11-loss was noted. The overall condition of the' piping and turning vanes was good.

Visual inspections were also performed on target tees and ' piping iri erosion suspect areas. The toes are located downstream of 2-LCV-6-31A, 2-LCV-6-9A, and 2-LCV-6-58A, and their purpose is to absorb the damage caused by condensate flashing to steam as it passes through the level control valves (LCVs). Piping failures had previously occurred on the unit 1, 1-1/2-inch drain piping immediately downstream of the LCVs. 'This piping on unit 1 was changed to 304 stainless steel (SS)' to prevent failures. Identical areas on unit 2 were also replaced with the 304 SS material. Inspection of'the 8 inch-diameter piping downstream of the aforementioned LCVs showed no O

evidence of wall degradation. The surfaces were covered with a dark-V colored, high-temperature magnetite oxide. The general condition of the

- target tees downstream of the LCVs was similar, and no' appreciable damage could be detected.

UT was performed on the Nos. 2 and 3 extraction steam lines to determine if 3

- wall loss was present. These lines were selected by Larry Boyd, of OE, because the thermodynamic conditions in this piping are more conducive to j

steam erosion damage than in the other extraction steam lines. The piping I

was grided in squares, and testing was performed in accordance with N-UT-26. The minimum wall acceptance criteria was calculated for.each grided area based on.the following design parameters.

Grid No.1. No. 3 Extraction

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1 Design Temperature 3880F Design Pressure 200 psig Nominal Pipe Diameter 20 Inches Material S, specification ASTM A106, Grade B, Standard Weight TVA Class" H

1 Calculated Minimum Wall Value 0.2026 (includes minimum 0.070 inch

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corrosion allowance)

Minimum Wall Value Detected 0.290 m.

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Grid No. 2, No. 3 Extraction Design Temperature

.388CF Design Pressure 200 psig-Nominal Pipe Diameter 20 inches Material Specification ASTM A106, Grade B,. Standard Weight TVA Class H

Calculated Minimum Wall Value 0.2026-(includes minimum 0.070 inch corrosion allowance)

- Minimum Wall Value Detected 0.370 l

Grid No. 3, No. 3 Extraction I

Design Temperature 388or Design Pressure 200 psig Nominal Pipe Diameter 28 inches Nominal Wall Thickness 0 375 inch Material Specification ASTM A155, KC70, Class 2

- TVA Class H

Calculated Minimum Wall Value O.2473 (includes. minimum 0.088 inch corrosion allowance)

Minimum Wall Value Detected 0 350 Grid No. 4 No. 3 Extract'on i

Design. Temperature 3880F

.q Design Pressure 200 psig

'M Nominal Pipe Diameter 28 inches Nominal Wall Thickness 0 375 inch Material Specification ASTM A155, KC70, Class 2 TVA Class-H Calculated Minimum Wall Value 0.2473 (includes minimum 0.088 inch corrosion allowanc'e).

Minimum Wall Value Detecte'd 0.460 Grid No. 5. No. 2 Extraction Design Temperature-4220F Design Pressure 300 psig n

Nominal' Pipe Diameter 14 inches

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Material Specification ASTM A106, Grade B, Standard Weight

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Calculated Minimum Wall Value 0.2089 (includes minimum 0.070 inch corrosion allowance)

Minimum Wall Value Detected 0.330 e

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Design Temperature.

422oF.

d Design Pressure 300 psig Nominal Pipe Diameter 14 inches

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j Material Specification ASTM A106, Grade B, Standard Weight -

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Calculated Minimum Wall Value 0.2089 (includes minimum 0.070 inch corrosion allowance)

Minimum Wall Value Detected 0 360

' Grid No. 7. No. 2 Extraction

! Design Temperature 4220F Design Pressure 300 reig

.Nominsi Pipe Diameter 20 inches Material' Specification ASTM A106, Grade B, Standard Weight TVA Class H

Calculated Minimum Wall Value 0.2684 (in'l~udes: minimum 0.070 inch.

c corrosion allowance)

Ministat Wall Value Detected 0.390 The UT data did. not show any areas of serious wall degradation; however, evidence of wall thinning was detected on the lower extrados portion.of the No. 3 extraction line elbow which.is connected to the crossaround piping.

Wall. thickness values in this area are 0.29 inch and 0 31 inch; which

.3-suggests that some erosion activity is occurring. The calculated minimum 6) wall-value for.this area and all other areas examined was exceeded by the detected minimum wall value; therefore, based on the results of the UT inspection, it,is concluded that severe steam erosion damage does.not presently exist in the Nos. 2 and 3 extraction steam piping.

Attached are.the UT results of the extraction steam piping, sketches

- showing grid locations, and photographs of the target tees that were visually inspected.

Conclusion and Recommendations I

The visual and UT inspections of the subject piping illustrated that some,

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erosion does exist; however, it does not presently jeopardize the system #

integrity. Since this problem will progress, it is recommended that a surveillance instruction (SI) be initiated to monitor the grids located on the Nos. 2 and 3 extraction steam piping during outages for wall' degradation. This SI should be implemented until a trend can be established.

Once the trend is established, a permanent ins'pection frequency may be determined and corrective actions may be recommended as required. An identical inspection program is also recommended for unit 1 to monitor potential problem areas.

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S01 860403 928 t*NITED 8TATES COVERNa!ENT Cin0TdHditMI TENNESSEE VALLEY AUTI!ORITY To P. R. Vallace, Plant Msnacer, ONP, POB-2. Sequoyah Nuclear Plant C. R. Brimer, Manager. Site Services. ONP, 58-2. Sequoyah Nuclear Plant APR 03 W86 i

N' T SEQUOYAH NUCLEAR PLANT - INSPECTION OF HP AND LP CROSSAROUND PIPING, MSR LOV DRAIN STARTUP VENTS, NO. 3 HEATER DRAIN PIPING AND NOS. 1, 2, AND 3 EXTRACTION STEAM PIPING Attached for your review is a metallurgical inspection report on the subject i

piping. The results from the inspections identified wall loss in the No. 2 i

estraction steam piping. The remaining piping had no appreciable loss. I am recommending that the following actions be taken to monitor the piping, j

1.

To identify further damage, an ultrasonic wall monitoring surveillance instruction (SI) should be written to monitor the extraction steam piping at each refueling outage. This action should be implemented by j

the Systems Engineering Section.

y i

2.

MI 14.4 and 14.5 should be revised to include the metallurgical p

inspection of the HP and LP crossaround piping. This action should be implemented by the Codes and Standards Section.

I 3.

As damaged piping is identified, the scope of the SI ehould be changed

~

i to include the damaged areas to monitor wall degradation rates.

This action should be implemented by the Systems Engineering Section.

Review methods of improving steam quality to minimize erosion corrosion 4

damage to extraction steam and carbon steel secondary system i

compone.ts. This action should be implemented by the Systems Engineering Section.

5.

Eva'luate operation with secondary system pH c.t higher levels E 9.0 through 9.2.

The maximum acceptable secondary system pH should be determined in the system. This action should be implemented by the Chemical Engineering Section.

e p

.gj C. R. Brimer WSW:DFC: Hl.P:HC Attachment l

cc (Attachment):

I R1M_S MR AN J2A-Cj u

R. M. Mooney. ONP, O&PS-3. Sequoyah g

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W. L. Williams ONP. 0F 03-4, foquoyah

. q' This was prepared,.rincipally by R. L. Phillips.

COORDINATION:

,.h R. M. Mooney/ Systems Engineering Section.

V. L. Williarr / Chemical Engineering Sequoyah I

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We have completed our ersluation of the grid locatione identified to es es the above ref sresced marierandes. The sisimum acceptable ea11 thiaming The vetees are based sa a deseweight sad pressure evalesties esly.

thiaming values aseeciated with each grid are listed helews Srid Pipe Sise Nominal t Min. t 1-1 29" 0.37 5 0.13 ggg h

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0.375 0.18 Thimming below the values shown should be f urther evaluated by CE.

Pteese contact T. J. Neams (6%6) if you beve an,que s t ion s.

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PJf_1 W_Akk THINNING INSPECTION DATA PLANT: [M UNIT:

A INSP. DATE:

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'.i sase.h. u UNITED STATES GOVERNMENT-l

.... Memorandum TENNESSEE VALLEY AUTHORITY-l 525

'870408 067 H. L. Abercrombie, Site Director. 04PS 4, Sequoyah Nuclear Plant a

to D. W.. Wilson, Project Engineer, Sequoyah Engineering Project, DSC-E, FROM Sequoyah Nuclear Plant l

DATE

. APR 081987

SUBJECT:

SEQUOYAH NUCLEAR PLANT UNIT 2 - WALL THINNING ASSESSMENT PROGRAM FINAL REPORT

References:

1.

My memorandum to you dated January 27, 1987, "Sequoyah Nuclear Plant Unita 1 and 2 - Preliminary Report on the Condensate-Feedwater Piping Inspection - Suspected Erosion-Corrosion Areas" (B25 870127 028)

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2.

C. R. Brimer's memorandum to L. M. Nobles dated March 6, 1987, "Sequoyah Nuclear Plant Unit 2 - Preliminary Evaluation of the Turbine Building Heat Cycle Piping" (B29 870306 001--copy attached)

Attached is the final report on the. unit 2 wall thinning inspection i

program. The inspection initially began with the condensate-feedwater system, but was expanded to include all heat cycle piping. The results identified localized damage on the feedwater piping and high pressure vent lines. The major damage of the feedwater piping was attributed to a situation caused by a 12-inch feedwater flow control valve (FCV) in a 16-inch line that introduces 9-inch diameter restrictions. This caused the fluid to accelerate to high velocity before discharging downstream and subsequently degrading the carbon steel 16-inch schedule 80 elbows, the 16-by 18-inch increasers, and the 18-inch diameter piping. The piping geometry downstream of valve 2-FCV-3-103 is. slightly different than that of the other three FCVs. An 18-inch 450 elbow is immediately downstream of the 16-by 18-inch increaser rather than th'e' 18-inch piping as identified of or 1 th d

oni ta con the dama as being minor. Major damage was obcerved in a 2 1/2-inch schedule 80 elbow on operating vent line "A." Vent lines "B"

& "C" adjacent to the damaged l

line did not show any significant thinning.

Minor damage was observed on a reducing elbow during the initial condensate-feedwater piping inspection; however, less than 3 percent thirming was l

found, and it appeared to be cavitation and/or an anomaly (see reference 1).

?

Ruv U.S. Savines Bonds Regularly on the Pavrall Sarinn Plan

i, 2'

H. L. Abercrombie APR 081987 SEQUOYAH NUCLEAR PLANT UNIT 2 - WALL THINNING ASSESSMENT PROGRAM FINAL REPORT As a result, I am recommending that the identified damaged 16-inch

.i elbows and 16-by 18-inch reducers be replaced with a more erosion-J corrosion resistant material, preferably 304 stainless steel. A carbon steel base metal repair should be performed on the localized areas showing minor wall thinning on the 18-inch 450 elbow and the 18-inch diameter piping. Also, I am recommending that the Materials Engineering Section maintain a surveillance instruction and a database to monitor those areas inspected for the remainder of the plant's life.

d vm fj g X H. Wilson CRB:RDE:RLP:HC Attachments oc (Attachments):

RIMS, SL 26 C-K (w/o reference)

, M. R. Harding, O&PS 4, Sequoyah (Attn:

L. M. McCormick)

R. E. Daniels, DSC-H, Sequoyah (Atta:

J. E. Pilgrim)

L. M. Nobles, P08-2, Sequoyah (Atta:

G. S. Boles)(w/o reference)

J.,H. Sullivan, SB-2, Sequo Principally Prepared By:

R. L. Phil11 tis d T. R. Woods f

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SEQUOYAll NUCLEAR PLANT (SQN) UNIT 2 WALL THINNINO ASSESSMENT PROORAM -

FINAL REPORT Introduction SQN recently initiated an inspection program to identify wall. thinning in its bulk single and dual phase flow systems. A preliminary inspection was performed on selected fittings to determine if ger.oric thinning was occurring on the condensate-feedwater and extraction steam piping. The condensate-feedwater inspection was performed in December 1986, and the extraction steam was inspected in October 1984 for unit 2 and May 1985 for unit 1.

The results did not reveal any significant thinning; however, the Materials Engineering Section evaluated technical information from the Surry Nuclear Plant event and decided 'to perform a detailed investigation of all the heat cycle piping. The systems that were evaluated were as follows:

Single Phase Dual Phase Condensate Extraction. Steam Feedwater Heater Drains & Vent Lines Turbine Drain & Vent Lines Each were evaluated based on flow velocity, operating pressure, operating temperature, and geometric configuration. The water chemistry and material compositions were fixed variables and were assumed as constants. From the evaluations, approximately 150 areas were identified as being the most susceptible. Of these 150 areas, 70 were targeted for examinations.

Scove The purpose of the assessment program was to determine the high suspect

\\

areas and then to inspect and monitor these areas for the entire plant I

life. Surry had seen major damage ~ after, a' approximately 76,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of

}

operation, whereas, SQN has experienced approximately 25,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.' The long-term monitoring initiated should prevent the occurrence of a catastrophic rupture. Thc assessment program would also make any corrective action rN emendations. The program should identify wall thinning due to cavitation (high pressure differential), pure erosion (high velocity), pure corrosion (low velocity), erosion-corrosion, and any anomalies that would violate the design minimum wall thickness.

Results Ultrasonic Testing Approximately 70 areas (Tables 1, 2, and 3) were examined by ultrasonic testing in accordance with N-UT-26. The results were compared to the manufacturer's minism wall thicknesses and tho' Division of Nuclear Engineering's design minimum wall thicknesses. Any deviations found below the manufacturer's minimum were noted, and those areas found below or rapidly approaching the design minimum were targeted for replacement.

s 4

ee a

4 The results identified two areas on the condensate-feedwater system with damage and one area on the high pressure operating vent lines.' Significant damage was found downstream of the 12-inch feedwater valves. This damage was found on the 16-inch elbows and the 16-by 18-inch increasers. The damage was measured to be 30 percent below the design minimum wall thickness in some localized areas on the 16-inch elbows and within 12 1

percent of the minimum design value on one of the increasers.. Each of the increasers showed varying degrees of wall thinning. Hinor damage was noted on the 18-inch 450 elbow downstream of the 16-by 18-inch increaser on the No. 4 feedwater line and the 18-inch piping ' downstream of the 16-by 18-inch increaser on the Nos.1, 2, and 3 feedwater lines. Significant damage was also identified on the "A" high pressure operating vent line.

Lines B and C adjacent to the damaged line showed some thinning, but nothing significant. Each line has been targeted for replacement.

Minor damage had been detected earlier during the condensate-feedwater piping inspection on a reducing elbow of the main feedwater pump discharge. The damage was less than 3 percent of the manufacturer's mini ma (see reference 1).

Visual Testing Approximately 11 elbows on the condensate-feedwater system were' visually j

inspected for wall thinning. The 10 elbows upstream and downstream of the No. 2 feedwater heater did not show signs of damage (see reference 1). The elbows, piping, and fitting immediately downstream of feedwater regulating valves showed varying degrees of wall thinning. There was a clear demarcation revealing exposed base metal and protective magnetite. These conditiona are very characteristic of erosion-corrosion damage (see reference 1).

Chemical Analysis Seven elbows were analyzed for beneficia1Jtrace elements to complement the visual and ultrasonio inspectims. Six ' elbows were randomly selected for analysis. All had beneficial alloys present, and the ultrasonic data did not identify thinning. The elbow downstream of 2-FCV-3-103 (No. 4 feedwater line) was analyzed, and no beneficial alloying element was found. This elbow had experienced significant wall degradation.

Discussion and Recommendations SQN wall thinning assessment program did not identify any widespread wall thinning as did Surry. The thinning was confined to isolated areas and attributed to the inherent design. The primary wall thinning mechanism that was $dentified was erosion-corrosion, both single and dual phase.

Figures 1 and 2 show the effects of velocity, pH, and residual beneficial elements. Figure 3 shows the effect of geometry. The operating parameters 9

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ti:

4 and the inspection confirmed the previous assumption that SQN has maintained good water chemistry, the primary cause of wall thinning. The Materials Engineering Section recommends the following corrective actions.

1 1.

The Materials Engineering Section shall establish a data base to maintain all inspection information.

2.

The Materials Engineering Section shall write surveillance instructions to monitor the suspect areas.

3 The Mechanical Maintenance section shall replace degraded fittings with a more erosion-corrosion resistant material, preferably 304 stainless steel. Also a carbon steel base metal repair should be performed on the localized areas showing minor wall thinning on the 18-inch 450 elbow and the 18-inch diameter piping.

RLP:DR 3/26/87 HC7083 01 a

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'g Tdl31 SEOUOYAN NUCLEAR PLANT WLL TMlWIM PR06AAN REFDDCE LIST 18NUR8ER DESCRIPT!GN ll!E SCH DE316N DES 16N MATDIAI.

PHASE PRESS.

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2 FN- -48 90 ADUCIll6 ELL nFPT4 $UCTION 16.000 M US 400 ASTN AI N GR.3 MTER 2-FW-5A 90 ROUCIM ELL WPT4 24.000 80 1985 600 ASTN A!04 ER.3 UATER IISCMR6E 2-FB-2 90 ROUCIM ELL WPT4 16.000 to 1985 600 ASTN A104 SR.3 NATER DISCMR6E t

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Ttblo 2 Page Ms.

2 IEOUOYM NUCLEAR PLMT WALL TH WING PR06RAN REFDDCE Ll'5T 18 NURIER DESCRIPTION

$1!E SCN M5!$N DESISM MTD!AL PHASE PRESS.

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24 6 90 ELL UP$TRN E V FCV-2-221 24.000 40 675 400 ASTN AIM ER.3 UATER 2-FW

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24 W-108 90 RDUCIM ELL NFPT-A SUCTION 16.000 40 675 400 ASTN AIM SR.3 MTR I

24W-11 90 ELL DuSTM RV FCV-2-128 20.000 40 675 400 ASTN AIM SR.3 MTER 2 4 12 90 ELL MITAN E V FCV-2-167 20.000 40 475 400 ASTN AI M SR.5 MTER 2 4 13 90 ELL DRITM Ef FCV-3-II, 24.000 10 1005 600 ASTN A106 GR.3 MTU NFFT3 24u-14 to ELL DusTM EV FCV-3-67, 24.000 le 1005 600 A87N AIM 6R.B MTR RFFT-A 24W-16 FM 3YPA88 TO COND UPSTM NCV 1.000 80 1005 600 ASTM A104 6R.3 UATR 3-70 424W-17 FEDMTR LOOP 4 90NNETEM OF !&.000 to 1995 600 ASTN A106 S.3 MTR FCVmM 245-II FIGHTR BYPA8S MA8G WF 0F 8.000 30 1985 600 ASTN AI M SR.3 UATR f

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't me eva u,iss.s.m J ' $/NITED STATES GOVERNMENT 1

. Memorandum TENNESSEE VALLEY AUTHORITY Ti25

'870127 0428

~

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'TO' H. L. Abercrombie, Site Dinctor, ONP, Osca-4, Wq'IiTy'ali"NicTiiF'Prant. ~- ~

--==

FROM

- D. W. Wilson, Project Engineer, Sequoyah Engineering Projet::t, DNE, DSC-E b

Sequoyah Nuclear Plant

'DATE JAN 2 71987-

SUBJECT:

SEQUOYAH-NUCLEAR PLANT UNITS 1 AND 2 - PRELIMINARY REPORT ON THE COND FEEDWATER PIPINO INSPECTION - SUSPECTED EROSION-CORROSION AREAS Attached for your review is the preliminary report of SQN condensate-

)

feedwater inspection. The results indicate that there is no wall thinning =

l

- due to erosion-corrosion. However, there may be minor (three-percent wall thinning) cavitation damage on the discharge piping of A and B feedwater pumps. The remaining wall in that area has not been reduced below the minimum design wall thickness. Appropriate surveillance instructions shall be written to monitor the suspect areas. The instruction will be written by Operations Engineering Services'-

metallurgical employees and is expected to be in place by June 30, 1987.

i The final report will include the results or ultrasonic examinations of the elbows downstream of A and B pump and will be issued the' week of I

February 6, 1987.

l

{

R D'. W. Wilson RBfD 0:RL Attachment

/

oc (Attachment):

./

4 JtIMS. SL 26 c-K

/

1 M. J. Burzynski, ONP, O&PS-4, Sequoyah 1

i J. C. Eey, DNE, DSC-E,' Sequoyah J. H. Sullivan, ONP, SB-2, Sequoyah B. M. Patterson, ONP, POB-2, Sequoyah (Attn:

E. L. Booker)

Principally Prepared By: Robert L. Phillips and Terry R. Woods, extension 6946 4

9 i

s L

HC7017.01

(

1 b_ere nan u s n a a_ ~ m w w w-

1 SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 - PRELIMINARY REPORT ON CONDENSATE-FEEDWATER PIPING INSPECTION - SUSPECTED EROSION-CORROSION AREAS

References:

1.

D. W. Wilson's memorandum to H. L. Abercrombie dated December 19, 1986, "Sequoyah Nuclear Plant Units 1 and 2 -

^

Inspection of Feedwater Piping' for Wall Loss"

~

(B25 861219 001) 2.

Report by P. Berge and F. Khan, of Electricity de France, dated May 1982, " Corrosion Erosion of Steels In High Temperature Water and Wet Steam" 3

EPRI NP 3944 report, " Erosion / Corrosion in Nuclear Plant Steam Piping; Causes and Inspection Program Guidelines"

Background

l On December 9,1986, Surry Station Nuclear Plant had a pipe rupture on the condensate-feedwater system that caused several fatalities. The rupture was caused by localized wall thinning at a pipe-to-elbow weld. The thinning mechanism was identified as erosion-corrosion (EC). Sequoyah Nuclear Plant (SQN) implemented a program to identify possible EC damage (see' reference 1).

The program was developed from technical information from Surry Station, INPO network, regional and resident NRC inspectors,. and information from references 2 and 3 EC is characterized by dissolution of protective magnetite film by a high temperature liquid stream in contact i

with steel surfaces. EC damage is normally found in elbows on the extrados (outer radius); however, it may also be seen on the intrados (inner radius). The phenomenon is usually observed in plain carbon and low alloyed steels at elevated temperatures. The following are factors influencing the EC mechanisms.

1.

pH and water and/or steam chemistry 2.

Material composition i

3 Flow path geometry 4.

Velocity 5.

Temperature Incorporating the' above factors and experience from Surry Statien, a

. temperature toundary of 300 to 400 degrees Fahrenheit was established for initial inspection. These areas were considered to have the highest probability of danage. The locations inspected are identified in figures 1, 2, and 3.

Surry and SQN both used ASTM A106 Orade a piping and fittings on the feedwater system. The plants also had similar operating parameters at the time of failure (i.e., water chemistry). The piping that failed had e

• S 9

m

,g l

t thinned from 0.500 inch- (nominal wall) to 0.060 inch.

SQN has schedule 40-piping (nominal wall.688 - minimum wall'.602).

The rupture at'Surry occurred at the feedwater pump section at a spool-piece and a 90-degree elbow.

SQN has a spool-piece and 45 degree elbow, which has a less severe -.. w;

. geometry factor. SGN's feedpump section elbow and spool-piece and.

the suspect areas adjacent to the A and a pumps were selected for l

l

~

inspection.

Objective

.The purpose of this report is to:

l i

' Identify possible thinning.

1.

2.

Determine if thinning had exceeded the minimum wall thickness.

i-3 Recommend corrective action.

4.

Write, as necessary, preventative maintenance (PM) and/or surveillance instructions (SI) to trend degradation.

The inspection plan employed ultrasonic (UT) and metallurgical inspection methods.

All UT was performed in accordance with TI-51,'N-UT-26.

The Inservice Inspection Group (ISI) performed the UT.

Tha metallurgical inspections were performed by Operations En6dneering Services- (CES),.

Welding and Metallurgy Section and utilized flashlights and visual?

tids.

Both inspection methods will identify EC damage as intermittent ID surface grooving and gouging. Visually, EC would be seen as intermittent removal of the black magnetite film exposing either bare metal or red hematite.

Results and Discussion 5C[

The UT was performed on the suspect pipe OD surfaces on selected elbow r

extradose and intradose utilizing 4-by 4-inch grids to idictify the examination. Each grid was evaluated, and the.naximem and :d.nimum readings

'j were recorded.

Gross differences in the high und low reading were' noted and evaluated by ISI and OES. Figures 4 through 28 show the typical inspection method, practices.

The piping and fittings were schedules 40 and 80.

Although the fittings do not have a standard nominal wall, they are manufstetured to a minimum thickness in accordance with ANSI B16.9,16.11, and 16.28.

The minimum for each size and schedule was calculated (see tables 1 and 2).

This minimum was used asa baseline. Readings below the minimum would indicate thinning.

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3-The UT data was uniform and consistent and indicated that there. was no thinning occurring as a result of EC, which would appear to be localized areas of non-uniform thinning. With one exception, there were no readings J.

below the minimue thickness established in accordance with ANSI

'a specifications. Wall thickness measurements taken on the discharge side of the feedwater pump on a 24-by 16-inch reducing elbow (Grid 2-FW-9) showed some evidence of Wall loss. This wall reduction is believed to have resulted from cavitation damage because cf the large pressure drop that exists at that location. Although three-percent wall reduction was noted, the minimum wall acceptance criteria for this fitting had not been violated, and this area will be monitored for wall reduction in the future.

The Division of Nuclear Engineering (DNE) had provided a design minimum acceptable wall thickness for the areas identified for the analysis (see tables 1 and 2). The inspections showed that no reading was below these values.

Metallurgical Inspection Metallurgical inspections were performed on A and C t' rains of units 1 and 2 number 2 feedwater heaters. The locations are shown on figure 16.

Both the inlet and discharge piping and fittings were inspected. The inlet piping had some superficial patterns on its wall because of direct impingement from the number 3 heater drain tank piping. No red hematite was observed on the ID, and the black magnetite fiLn was intact (see location 3, figure 16).

At. location _2, no red hematite or exposed base metal was observed.

On the discharge piping, the results were similar.,

. Also, there was a backing-ring that had been pushed into the flow piath

• during original installation. It showed no signs of wear and was covered with the protective magnetite film, even though it was in a severe environment.

Discussion UT and metallurgical inspections indicated that no EC damage or significant thinning by _other means was detected, although SQN has conducive feedwater' piping conditions. However, SQN has maintained good feedwater chemistry, which lessens the probability of failure and/or EC damage. The history of the feedwater chemistry at Surry Station is unknown. Previous inspections on the number 3 heater urain tank, the steam generator feedring header, and the feedring tee did not reveal service-induced damage. EC damage was observed on the feedring J-tubss.

(The J-tucas wera A106 ' Grade'B steel, but the velocities were as high 31 ft/sec.) Velocity of the 24-and 30-inch headers and fittings were 12 ft/sec and 14 ft/sec respectively (see table 3). The propensity of the EC decreases with a decrease in velocity.

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Conclusions and Recomc:endatfens The test data and inspection results it.dicated that EC dama6e had not occurred in the areas t;xamined. The_sslect/td areas were identified as the

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. highest probability areas. Ecwever, there way be other thinning mechanisms accurrin6,' i.e., cavitation. The lokest" readings were found on the discharge side of the feedwate.c pump on 24-by 16-inch reducing elbows.

None of these readings.wcre below taa desi6n e nimum wall thibkness specified.by DNE. The elbows further downstream of the A and B pumps will-be examined and included'in th( final repor't. The. piping upstream of the

- i

. pumps is ceroptable but should be monitored 'cy an SI each refueling outage. ' Feedwater pH should be optimized to the highest pH attainable to j

minimize Yhe, potential ^for EC damage throughou.t the balance of the plant

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carbon st, eel, system. A study to optimite. the SQN pH has been' initiated by l

the Chemical' Engineering Unit.

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TABLE 3 Original. mass flow rate:

15,705,450 lbs/hr

'.._'s Conversion factors

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1 hr

= 3600 see 3

I ft 7.48 gallons 1 gallon s 8 337 lbs

~~

Calculation 3

3 15,705,450 lbs x

1 he x

1 ft x

1 gal

= 69 96 ft/ 3,c 1 hr 3600 sec 7 48 gal 8,337 lbs Yelocity :

Mass flow rate Pipe radius = 15 inches or 1.25 feet cross sectional area 3

Velocity = 69 96 ft /see 14.25 ft/seo for 30-inch 2

4 909 ft Assume mass flow rate equally divides into two 24-inch pipes.

New mass flow rate = 7,852,725 lbs/hr Pipe radius 12 inches of 1 foot 3

3 7,852,725 lbs x

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34.979 ft /see hr 3600 sec 7.48 gal 8,337 lbs 3

Yelocity = 34.979 ft /sec =

11.13 ft/sec for each 24-ined pipe 2

3 14 ft Reviewed by DNE

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' Fibre 4 Grid Nc. 1-FW-1 24-inch, schedule 40 elbow (upstream of B pump).

Minimum wall thickness in accordance with ANSI B16 9 is 0.602-inch.

• Minimum wall thickness measured by UT is 0.740-inch.

8 Readings were uniform. No erosion-corrosion damage detected.

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.gigure5 Grid Number 1-W-2

'2k-inch, schedule 40 spool-piece and 45-degree elbow (upstream of B pump). Minimum wall thickness in accordance with ANSI B16 9 is. 0.602-inch.

• Minimum wall thickness measured by UT is 0.620-inch.

• Readings were uniform. No erosion-corrosion detected.

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.c Grid Number 1-FW-3 30-inch, schedule 40 header (upstream et a pump).

Minimum wall thickness in accordance with ANSI B16 9 is 0.656-inch.

• Minimum wali thickness measured by UT is 0.760-inch.

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Figure 7 i-Grid Number 1-FW-4 24-by 16-inch, schedule 80 reducing elbow (discharge side of B pump). Minimum wall thickness in accordance with ANSI B16 9 is 0 740 inch

• Minimum wall thickness measured by UT is 0.830-inch.

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Figure 8 Grid Number 1-FW-5

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l 24-inch, schedule 40 elbow and 24-by 16-inch, schedule 40 reducing elbow (upstream B pump).

Minimum wall thickness in accordance with ' ANSI B16.9 is:

24-inch elbow - 0.602-inch and 24-by 16-inch reducing elbow - 0.4375-inch.

• Minimum wall thickness measured by UT is 0 740-inch.

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7 Grid Number 1-FW-6

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Minimum wall thickness in accordance with. ANSI B16 9 is 0.602-inch.

• Minimum wall thickness measured by UT is 0 740-inch.

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Grid Number 1-FW-7 24-inch, schedule 40 header and 45-degree elbow (upstream A pump). Minimum wall thickness in accordance with ANSI B16 9 is 0.602-inch.

8 Minimum wall thickness measured by UT is 0.660-inch.

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figure 11 Grid Number 1-W-8 24-inch, schedule 40 elbow (upstream A pump).

Minimum wall thickness in accordance with ANSI B16.9 is 0.602-inch.

• Minimum wall thickness measured by UT is 0 740-inch.

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Minimum wall thickness in acccedance with ANSI B16 9 is 0 740-inch.

• Minimum.rall thickness measured by UT is 1.100-inch.

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.F.igure 13 Grid Number 1-FW-10

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l 24-in'ch, schedule 40 elbow and 24-by 16-inch, schedule 40 reducing elbow (suction side A pump). Minimum wall thickness in accordance with ANSI B16 9 is:

24-inch elbow - 0.602-inch and 24-by 16-inch reducing elbow - 0.4375-inch.

s Minimum wall thickness measured by UT is 0 740-inch.

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"Eigure 14 Grid No. 1-FW-11 20-inch, schedule 40 elbow (upstream A train, No. 2 feedwater heaters). Minimum wall thickness in accordance with ANSI B16 9 is 0 520-inch.

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• Minimum wall thickness measured by UT is 0 560-inch.

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Figure 15 a

Grid number 1-W-12

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20-inch, schedule 40 elbow (upstream C train No. 2 feedwater heater). Minimum wall thickness in accordance with ANSI B16 9 is 0.520-inch.

' Minimum wall thickness measured by UT is 0.580-inch.

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.r Unit 1 C-Train No. 2 Feedwater Heaters Metallurgical inspections were performed at numbered locations. Results are as follows-1.

No degradation of adhering magnetite film.

Backing-ring protruded into fluid stream. No wear was observed on the backing-ring.

No degradation of adhering magnetite film.

3 superficial impingement from inlet stream. No degradation observed.

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' Figure 17 i

Grid No. 2-FW-1 24-inch, schedule 40 elbow (upstream of B pump).

Minimum wall thickness in accordance with ANSI B16.9 is 0.602-inch.

• Minimum wall thickness measured by UT is 0.680-inch

• Readings were uniform. No erosion-corrosion damage detected.

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Figure 18 Grid No. 2-FW-2 24-inch, schedule 40 spool-piece and 45-degree elbow (upstream of B pump). Minimum wall thickness in accordance with ANSI B16 9 is 0.602-inch.

• Minidum wall thickness measured by UT is 0.620-inch.

• Readings were uniform. No erosion-corrosion detected.

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l F'igure 19 Grid No. 2-FW-3 30-inch, schedule 40 header (upstream or B pump).

Minimum wall thickness in accordance with ANSI B16.9 is 0.656-inch.

• Hinimum wall thickness measured by UT is 0.760-inch.

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24-inch, schedule 40 elbow and 24-by 16-inch, schedule 40

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reducing elbow (upstream b pump). Minimu:n wall thickness j

in accordance with ANSI B16.9 is: 24-inch elbow - 0.602-inch and 24-by 16-inch reducing elbow - 0.4375-inch.

e Minimum wall thickness measured by UT is 0 550-inch.

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" Figure 21 Grid No. 2-FW-5 24-by 16-inch, schedule 80 reducing elbow (discharge side of B pump). Minimum wall thickness in accordance with

, ANSI B16.9 is 0.740-inch.

• Minimum wall thickness measured by UT is 0.680-inch.

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24-inch, schedule 40 elbow (upstream A pump).

Minimum Wall thickness in accordance with A!ISI B16.9 is 0.602-inch.

• Minimum wall thickness measured by UT is 0.710-inch

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Figure 23 i

Grid No. 2-FW-7 24-inch, schedule 40 header and 45-degree elbow (upstream A pump). Minimum wall thickness in accordance with ANSI B16.9 is 0.602-inch.

• Minimum wall thickness measured oy UT is 0.610-inch.

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m Tigure 24 Grid No. 2-FW-8 24-inch, schedule 40 elbow (upstream A pump).

Minimum wall thickness in accordance with ANSI B16.9 is 0.602-inch.

8 Minimum wall thickness mea.sured by UT is 0.700-inch.

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' Figure 25

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Grid No. 2-FW-9 24-by 16-inch, schedule 80 reducing elbow i

(discharge A pum;}. Minimum wall thickness in accordance with f.NSI B16 9 is 0.740-inc!4.

1. Minimum wall thickness measured by UT is 0.720-inch'.

8 Suspect minor cavitation damsgu in Grid,I 14

~(see photo above) i

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Grid No. 2-FW-10 24-inch, schedule 40 elbow and 24-by 16-inch, schedule 40 recucing elbow (suction side A pump).

~

Minimum wall thickness in accordance with ANSI B16.9 is: 24-inch elbow - 0.602-inch and 24-by 16-inch reducing elbow - 0 3475-inch.

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• Minimum wall thicknesa measured by UT is 0.670-inch.

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Grid No. 2-N-11 20-inch, schedule 40 elbow (upstream A train, No. 2 feedwater heaters). Minimum wall thickness in accordance with ANSI B16.9 is 0 520-inch.

• Minimum wall thickness measured by UT is 0.600-inch.

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• Figure 2B Grid No. 2-FW-12 20-inch, schedule 40 elbow (upstream C train No. 2 feedwater heater). Minimum wall thickness in accordance with ANSI B16.9 is 0.520-inch.

• Minimum wall trickness measured by UT is 0.580-inch.

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Z'E____I~_~_'_i_'_'E____~~~'.'_'

... [

P.IZI tfALk THINNING INSPECTION DATA INSP. DATE: Al" k

/VS PLANT:

N ON UNIT:

I EM Sh-

1. o 6.

o s - A tr R VIA L ;. e -

STsTEM:

MTL. CLASS./REFEFENCE THICKNESS: Ia-e-kof Tll, f LOCATIONS INSPECTEDi

_G PJ's "Tolle f d dfacL.A c[.s

,1, f

PURPOSE OF INSPECTION:

8 rl d /lr (

1m [;,

u m)I 4dlMffer FOR WALL THINNING 7 We INSPECTION DETAILS:

Ice Ee,4, cf METHOD:

UT INSTRUMENTS:

USL 29 UT GRIDS:

G-

1. %J b ;3 1 Yec Rel ort INSPECTION RESULTS:

4 CORRECTIVE ACTIONS:

d/A REPAIU/ REPLACEMENT 7 MATERIALS CHANGES 7 FAILURE ANALYSIS 7 DOCUMENTATION: Alee b T H M '.ller Y' ld I (oMe.

Ala Ed M3T, 8

(L2.aT5'or2.1 H J)

)

1

7 4cc

.e....,-

' (lNITED STATES GOVERNMENT h29 860.57,.3 990 Memorandum TsnutssEs vAttzr Aurnontry j.

TO' W. - T. Cottle, Site Director, NUC PR, t'atts Bar Nucimr Plant J. H. Miller, Chief !!echanical Branch,1270 CST 2-C FROM DATE

.i\\; *., 305 StfBJECT: WATTS BAR NUCLEAR PLANT UNIT 1 - BASELINE INSPECTION OF EXTRACTION STEAM PIPING AND HIGH-PRESSURE MOISTURE. SEPARATOR REHEATER VENT LINES Attention:

C. Nelson i

Attached for your information is a copy of the inspection report pertaining to the subj ect piping.

J. H. Miller GJP:JHF:Th:JLR' Attachme Y cc (Attachment):

NUC PR Rais, 1520 CST 2-C J. Inger,'NUC PR, Watts Bar i

This was prepared principally by T. R. Woods.

All26B.JR 1

~

Buy l!.S. Sa::ings Bonds Regularly on lite Payroll Sa::tn;;s P!an

__^ ' I I __ __-___

~ ~ ~ ~ "