NUREG!CR-5117

PNL-6446

Steam Generator Tube

Integrity Program/Steam

Generator Group Project

Final Project Summary Report

Prepared by R. J. Kurtz, R. A. Clark, E. R. Bradley, W. M. Bowen,

P. G. Doctor, R. H. Ferris, F. A. Simonen

Pacific Northwest Laboratory

Operated by

Battelle Memorial Institute

Prepared for

U.S. Nuclear Regulatory Commission
(

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NUREG/CR-5117

PNL-6446

R5

Steam Generator Tube

Integrity Program/Steam

Generator Group Project

Final Project Summary Report

Manuscript Completed: April 1990

Date Published: May 1990

Prepared by

R. J. Kurtz, R. A. Clark, E. R. Bradley, W. M. Bowen,

P. G. Doctor, R. H. Ferris, F. A. Simonen

Pacific Northwest Laboratory

Richland, WA 99352

Prepared for

Division of Engineering

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

NRC FIN B2097
K

For sale by the Superintendent of Documents, U.S. Government Printing Office

Washington, D.C. 20402
ABSTRACT

The Steam Generator Tube Integrity Program/Steam Generator, Group Project

was a three-phase program conducted for the U.S. Nuclear Regulatory Commission

(NRC) by Pacific Northwest Laboratory.(a) The main goal of the program was

to provide the NRC with validated information on the reliability of

nondestructive examination techniques to detect and size flaws in steam

generator tubing and to determine the remaining integrity of service-degraded

tubing. The information was used to determine the effectiveness of NRC

Regulatory Guides 1.83, Rev. 1, and 1.121 to 1) define the frequency, extent

and procedure for conducting nondestructive inservice inspections of steam

generator tubing, and 2) defin' plugging limits of service-degraded tubing

under normal operating and accident loading conditions. The program was

performed in three phases. The first phase involved burst and collapse tests

and single-frequency eddy-current (EC) examinations of typical steam generator

tubing with precision machined flaws. The goal of Phase I was to develop

empirical models of remaining tube integrity as a function of flaw type and

size, and to determine the capability of EC inspection methods to detect and

size tube degradation. In Phase II, a smaller number of specimens with the

same flaw types were investigated, but tube specimens were degraded by chemical

means rather than machining methods. This approach was used to better simulate

the irregular geometry of service-induced degradation. In the final phase of

the program, the retired-from-service Surry 2A Steam Generator was used as a

test bed to investigate the reliability of inservice EC inspection equipment,

personnel, and procedures, and as a source of service-degraded tubes for

further validating the empirical equations of remaining tube integrity. In

addition, the generator was also used to study the effects of primary-side

chemical decontamination, the effectiveness of secondary-side visual

examinations, characterizationý:of tubesheet crevice corrosion products and

sludge pile composition, and demonstrations of tube unplugging and repair

techniques. This portion of the program also included participation by three

foreign countries.

This report summarizes the findings of more than eleven years of research

activity on steam generator tube integrity and inspection issues. The results

of the Phase I pressure tests on mechanically-flawed steam generator tubing

are presented. In addition, the laboratory EC sizing results on those tubes

are summarized. A discussion of Phase II pressure test results on

chemically-degraded tubes is given, along with a brief overview of leak-rate

data for tubes with laboratory-produced axial or circumferential

stress-corrosion cracks (SCC). Comparisons with a simple analytical leak

rate model are discussed. Laboratory EC estimates of flaw size in Phase II

specimens are described. To supplement the laboratory EC data and obtain an

estimate of EC reliability to detect and size SCC, results of a mini-round

robin involving several firms that routinely perform field inservice

inspections are presented.

(a) Pacific Northwest Laboratory is operated for the U.S. Department of

Energy by Battelle Memorial Institute under Contract DE-ACO6-76RLO 1830.

iii
A major portion of the report is devoted to summarizing and integrating

the results from the more than six years of research on the Surry 2A Steam

Generator. A brief description of the acquisition, transportation and

placement of the steam generator into a specially built examination facility

is given. Prior to performing extensive nondestructive examination (NDE) of

the generator tubing, dilute chemical decontamination of the channel head and

removal of plugs placed during service was performed. To determine a

post-service baseline condition of the generator, two nearly 10.0%

multifrequency EC inspections were performed along with visual inspections

from the secondary side. From these inspections, a subset of 320 tubes was

selected for round robin inspection and development of the data base on EC

inspection reliability. A description of the four round robins conducted is

provided. A summary of the methods used to remove more than 550 tube segments

to validate the round robin inspection data is given. Results from the

metallurgical and visual examinations of these specimens are discussed. Burst

tests of removed-from-generator specimens with pitting/wastage-type

degradation are presented and correlated° with the empirical models of

remaining tube integrity. Statistical analyses of the combined metallurgical

and EC data to determine probability of detection and sizing accuracy are

reported along with a discussion of the factors which influenced the results.

Analyses and Monte Carlo simulations used to evaluate and compare various

sampling plans for inservice .inspection are described. Finally,

recommendations for improved inservice inspection and maintenance of steam

generator tube integrity are given.

iv
SUMMARY

In 1976, the U.S. Nuclear Regulatory Commission (NRC) initiated the Steam

Generator Tube Integrity Program (SGTIP) in response to an Atomic Safety and

Licensing Board inquiry regarding the existence of NRC data on the integrity

of service-degraded steam generator tubes. At that time the available data

were limited in quantity and had been generated solely by industry. The NRC

authorized the Pacific Northwest Laboratory (PNL) to perform a research

program to develop validated information on the reliability and effectiveness

of inservice inspection techniques to detect and size tube degradation and

the margin-to-failure of service degraded tubes under normal operating and

accident loading conditions.

The program was conducted in three phases. The first phase involved burst

and collapse tests, and single-frequency EC examinations of typical steam

generator tubing with machined flaws. Three flaw types were utilized:

uniform wall thinning, elliptical wastage and an axial crack-type flaw that

was simulated by an electrodischarge-machined (EDM) slot. These flaw

geometries were selected because they simulated the known or expected flaws

in pressurized water reactor (PWR) steam generators. Burst and collapse tests

were conducted in simulated PWR steam generator chemical and thermal

environments in a high-pressure autoclave assembly.

Due to the high burst pressure of unflawed steam generator tubing, even

tubing with fairly large flaws had burst pressures higher than could occur

during the worst credible burst mode accident. Typically, tubes with flaws

less than 80% through-wall had burst pressures higher than would be experienced

during such accident conditions. All of the collapse pressures of the flawed

tubing specimens were higher than would occur during the worst credible'

collapse mode accident.

The laboratory EC test results showed that single-frequency EC techniques

did not accurately size many of the machined flaw geometries. However, it was

also shown that the current plugging practices are conservative from the

viewpoint of margins-to-failure for the machined flaws studied.

Phase II of the program continued the tube integrity and EC examination

work of Phase I, but tube specimens were flawed by chemical means rather than

machining methods. This approach was chosen to better simulate service-induced

flaw geometries and thereby validate the equations developed in Phase I.

Uniform-thinning and elliptical-wastage specimens were produced by electroetching techniques, and stress-corrosion-crack-type degradation was

created by autoclaving mill-annealed tubing in a water environment with

impurity additions.

The Phase II burst and collapse tests were also performed in a simulated

PWR steam generator environment. Phase II failure pressures showed the same

general trends as the Phase I results with similar data scatter. Burst

pressures of SCC-flawed tubes were about 10% higher, on the average, than

those measured from the EDM slot flaw simulations. On the other hand, burst

pressures of uniform-thinning andelliptical-wastage flaws were less than 10%

v
lower, on the average, than those predicted from the Phase I empirical

relationships. Thus, additional conservatism for evaluation of

uniform-thinning and elliptical-wastage type flaws from the Phase I

relationships may be justified.

A review of the available leak-rate data for SCC-flawed steam generator

tubes subjected to normal operating and accident pressure differentials

indicated that the measured leak rates were highly variable when compared to

analytical predictions. Predicted leak rates were, in some cases, ten times

greater than measured.. Nevertheless, most (but not all) of the tubes leaked

at detectable rates for normal operating conditions and none of the tubes burst

at main-steam-line-break accident loadings. The data suggest that a

substantial level of conservatism should be applied to predictions of leakage

that are used for leak-before-break evaluations, and that such conservatism

would also appear appropriate for establishing leak detection limits for

detection systems.

Laboratory EC measurements of SCC depths indicated that this type of flaw

was, on the average, undersized, with the data displaying a great deal of

scatter. Elliptical-wastage and uniform-thinning flaws were more accurately

sized than SCC, with less scatter observed. Flaw sizing conservatism increased

as the amount of material removed, by the degradation mechanism increased.

Thus, uniform-thinning specimens were the most conservatively sized flaw

geometry.

The Phase I and iI.single-frequency EC results discussed above should be

considered only as a rough indicator of EC capability, since 1) the tests

were conducted under non-blind.laboratory conditions and 2) the test specimens

all had idealized'flaws and.no complicating factors to simulate real-world

conditions, Such as copper deposits, denting, support plates, and antivibration

bars. In addition, the tests were performed so that inspectors were aware of

the type of degradation being examined. The Phase III Surry investigation

provided a unique opportunity to gather NDE reliability information on tubes

with service-induced degradation' and extraneous effects under simulated field

conditions using multifrequency EC equipment. Thus, the information developed

more closely represents the reliability of actual inservice EC inspections

rather than a.measure of the capability of the equipment to detect and size

degradation under idea] conditions.

Phase III of the SGTIP because of its ambitious scope was organized as the

Steam Generator Group Project (SGGP). In this phase the retired-from-service

Surry 2A Steam Generator was used to 1) provide additional validation of the

tube integrity equations through burst testing of service-degraded specimens,

and 2) provide an extensive data base on the reliability of conventional and

near-term field practice EC inspection techniques to detect, characterize and

size degradation in an actual steam generator under simulated field

conditions. The SGGP was performed for the NRC with additional participation

by the Electric Power Research Institute (EPRI) and consortia from France,

Italy and Japan 'that included both industry 'and government organizations.

The Surry 2A steam generator'was transported by barge to the Hanford

Reservation near Richland, Washington. It was stored while the Steam

vi
Generator Examination Facility (SGEF) was completed. The SGEF was specifically

designed and constructed to house a removed-from-service steam generator in

its normal operating position. After completion of the SGEF, the generator

was placed into the facility by lowering it through a removable roof panel.

Radiological mapping of the generator and facility was then conducted. A

visual examination of the secondary side through shell penetrations originally

made at the Surry site and additional penetrations made at Hanford. then

followed. These examinations confirmed that the generator condition had not

significantly changed during shipment to Hanford. The secondary-side

examinations proved useful for qualitatively identifying conditions that could

lead to loss of steam generator integrity such as sludge accumulation at the

top of tube sheet or corrosion in the tube-to-tube support-plate crevice.

However, these observations did not provide quantitative information on

remaining tube integrity, except in cases of gross tube damage.

Following generator installation in the SGEF, the generator was prepared

for extensive nondestructive characterization from the primary side. First,

the channel head was decontaminated, using a different dilute chemical

technique on each side (hot-leg and cold-leg). This resulted in an

approximately 10-fold decrease in the radiation field in the channel head,

with no significant corrosive effects observed from the decontamination

solutions on the channel-head materials. Personnel dosimetry monitoring of

channel-head workers indicated that exposure to the eyes was the limiting

exposure. After this, 969 explosively placed tube plugs, used to remove tubes

from service that were degraded or possibly subject to degradation, were

machined out. This was performed to permit access to these tubes for

nondestructive inspection.

To characterize the number, type and distribution of tubing defects,

100% of the accessible tubes were examined using multifrequency EC techniques.

Two separate examinations were performed using field-experienced teams, one

using Zetec personnel and MIZ-12 equipment and the other Intercontrole

personnel and IC3FA equipment. The intent of these examinations was to

determine as best possible the overall condition of the steam generator, and

allow selection of a subset of tubes for more detailed round robin studies.

Following completion of the base-line inspections, a subset of 320 tubes

was selected for additional NDE inspection and analysis. Four separate but

related round robin exercises were conducted. The initial round robin

involved data acquisition and analysis by five teams utilizing the Zetec

MIZ-12 system and DDA-4 analyzer. This was followed by a nine team data

analysis round robin, with all teams analyzing an identical set of Zetec

MIZ-12 eddy-current signals. An equivalent round robin was also held in

Europe, with teams from Italy and France analyzing signals acquired with

Intercontrole IC3FA equipment. An advanced/alternate techniques round robin

consisting of NDE inspections with equipment or analysis procedures

significantly different from those used in either the base-line or other round

robin inspections was also performed. The data from the various examinations

were analyzed to establish the variability in defect detection and sizing.

This information was used to determine locations for removal of specimens for

destructive metallographic validation.

vii
To validate the NDE round robin results and determine the condition of the

Inconel 600 (a) tubing, more than 550 tube-segments were selected and removed

from the generator for metallurgical examinations and burst testing. Emphasis

was given to selecting specimens where EC indications had been reported.

However, specimens without EC defect indications were also removed to verify

the lack of defects or to establish the condition of the tubing at specific

locations. Specimens from all regions of the generator were removed and

examined, although the majority of the defect indications were reported near

the top of the tube sheet on the hot-leg side of the generator.

Pitting and wastage were the predominant tube defects found in the

specimens examined. This type of tube degradation was identified in the sludge

pile region above the top of the tube sheet (TTS), within the tube-to-tube

support-plate (TSP) crevice, and to a lesser extent at antivibration bar (AVB)

contact areas. 'The most severe pitting/wastage degradation was located in

the region 0 to 2 in. above the hot-leg TTS where wall-losses ranged up to

87%. Wide variations in the. dstribution and depth of local degraded areas

were observed both axially and circumferentially within the corroded region

of the hot-leg TTS specimens. These variations in defect distributions appear

to be a major factor in the variability of the EC depth estimates. In

general, wall-loss from pitting/wastage type degradation in specimens from

other regions of the generator was less than 20%.

Intergranular stresscorrosion cracking was also identified by visual

and metallurgical'examinations at specific locations within the generator.

Cracking was observed at the apex of Row 1 and Row 2 U-bends which was

attributed to stresses produced by inward movement of the U-bend legs from

corrosion of the seventh (uppermost) TSP. Denting at the TSP intersections

also produced cracking intubes removed from the hot-leg side of the generator.

Axial cracks initiated on the inner surface were found in specimens with

calculated strains as low as 10%. Calculated strains greater than 60% were

required to produce cracking from the outer surface.

Burst testing showed that pitting/wastage type defects did not appreciably

degrade tube strength. Comparison of the data with the empirical relationships

of tube integrity developed in Phase I indicates these relationships adequately

predict tube margin-to-failure. Defect length was found to be an important

consideration in proper defect evaluation. The EC estimates of pitting/

wastage depths were used to estimate remaining tube integrity. Since the EC

inspection teams did not report defect length, the reported defect depths in

combination with a defect length greater than or equal to one tube diameter

was assumed. By assuming a long axial length, a conservative estimate of

remaining tube integrity was obtained even when EC significantly undersized

the defect depth.

The probability of detection (POD) depends on the location and severity

of the defects. The POD for pitting/wastage type defects at the TTS increased

with wall-loss and approached 0.9 for defects with greater than 40% through-

(a) Inconel is a registered trademark of INCO Alloys International

(Huntington Alloys), Huntington, West Virginia.

viii
wall penetrations. Conversely, signals due to denting at the TSP intersections

interfered with the EC signals from cracks and the POD was near zero for these

locations. Insufficient numbers of other defect types and locations where EC

inspections were made precluded additional POD evaluations.

Wide variations in the reported EC depth estimates were observed between

specimens with similar wall-loss and also within the same specimen for data

from different inspection teams. The team-to-team variations for a given

specimen appear to result from differences in analysis procedures or the

analyst's interpretation of the complex EC patterns. Defect morphology and

distribution within the corroded region was considered the major cause for

the variations between specimens with similar wall-loss. However, dents and

surface deposits near the defects also contributed to the sizing variations.

In general, EC tended to undersize the pitting/wastage type defects, especially

for the severely-degraded specimens. Improved sizing accuracy was noted for

one team that employed special frequency mixes to enhance the signal-to-noise

ratio by suppression of signals due to denting, copper deposits and support

plates. Also, ultrasonic and rotating EC probes were successfully used to

augment conventional EC/bobbin-coil data to obtain improved sizing of

pitting/wastage.

Results from a mini-round robin performed with laboratory SCC-flawed tubes

showed that sizing depth and length were highly variable. Assessments of

remaining tube integrity based on EC estimated flaw dimensions and burst test

results were conservative, except for a few cases in which the level of

conservatism was somewhat less than the factor of safety of three currently

specified for the normal operating condition. Further, the best average

probability of detecting SCC for all inspection techniques was 0.63,

indicating that crack detection was difficult.

Analysis of various candidate inservice inspection sampling plans

indicated that even with 100% inspection most teams that inspected the steam

generator would, on the average, detect and plug about 65% of the defective

tubes (taken as >75% through-wall degradation). In contrast, the best

performing team could be expected to identify about 95% of the defective tubes.

Analytical studies and Monte Carlo simulations demonstrated that a 40%

systematic sequential sampling plan was almost as effective as 100% inspection,

assuming some clustering of degraded tubes. The simulation work also showed

that all systematic sequential sampling schemes examined (20%, 40%, and 100%)

were equally effective when the defective tubes are surrounded by large numbers

of degraded and defective tubes.

ix

ACKNOWLEDGMENTS

The Steam Generator Tube Integrity Program was a large and complex

undertaking that involved literally hundreds of people over its duration.

Many of the tasks performed had never been attempted before and required

extraordinary dedication and support from many people. Although'it is not

possible to properly express our gratitude to the many people that worked on

this project, we would like to acknowledge the contributions of those key

people without whose efforts this project would never have been successfully

completed.

We wish to express our profound appreciation to:

U.S. Nuclear Regulatory Commission

J. Muscara

C. Y. Cheng

B. D. Liaw

L. Abramson

PNL - Battelle

R. P. Allen

R. L. Bickford

A. S. Birks

J. F. Edwards

L. K. Fetrow

M. S. Goody

A. S. Jones

M. Lewis

G. J. Posakony

W. D. Reece

E. B. Schwenk

K. R. Wheeler

Electric Power Research Institute

C. S. Welty

J. P. N. Paine

J. A. Mundis

T. Oldberg

Commissariat a L'Energie Atomique

J. L. Campan

Framatome

K. Smith

J. Culembourg

J. Brunet

Xi
Ansaldo Componenti S.p.A.

R. De Santis

T. Badino

R. Gammino

Ansaldo S.p.A

F. Magris

M. Barberis

Energia Nuclear Ed Energie Alternativa

M. Cavaggioni

M. Brocco

F. Imperiali

Ente Nazionale Per L'Energia Elettrica

V. Regis

R. Pascali

G. Maciga

F. Bregani

Nuclear Power Engineering Test Center

M. Oishi

K. Ohno

H. Nakata

University of Tokyo

Y. Togo

H. Madarame

Mitsubishi Heavy Industries

0. Takaba

VEPCO

H. S. McKay

xii
CONTENTS

ABSTRACT ................................................................ iii

SUMMARY ................................................................. v

ACKNOWLEDGMENTS ......................................................... xi

1.0 PHASE I SUMMARY .................................................... 1.1

1.1 TUBE INTEGRITY TESTS .......................................... 1.1

1.2 SINGLE-FREQUENCY EDDY-CURRENT RESULTS ........................ 1.10

2.0 PHASE II SUMMARY ................................................ 2.1

2.1 TUBE INTEGRITY TESTS ....................................... .2.1

2.2 LEAK-RATE TESTS .............................................. 2.3

2.3 SINGLE-FREQUENCY EDDY-CURRENT RESULTS ......................... 2.7

3.0 PHASE III SUMMARY .................................................. 3.1

3.1 SGGP OBJECTIVES AND ORGANIZATION ............................. 3.1

3.2 THE SURRY STEAM GENERATOR .................................. 3.2

3.2.1 Generator History ...................................... 3.2

3.2.2 Generator Acquisition and Transport ................... 3.3

3.2.3 Steam Generator Examination Facility ................... 3.4

3.3 PREPARATION OF THE GENERATOR FOR RESEARCH ..................... 3.7

3.3.1 Health Physics, ALARA Radiation Control ................ 3.7

3.3.2 Channel-Head Decontamination ......................... 3.12

3.3.3 Tube Unplugging ........................................ 3.15

3.4 NONDESTRUCTIVE INSPECTIONS ............ ......................... 3.21

3.4.1 Secondary-Side Examination .......................... 3.21

3.4.2 Base-Line Eddy-Current Inspection ...................... 3.35

3.4.3 Round Robin Tube Selection ............................. 3.37

3.4.4 Round Robin Inspections ............................. 3.38

xiii
3.5 SPECIMEN REMOVALS AND NDE VALIDATION ..........................

3.5.1 Specimen Selection .....................................

3.5.2 Specimen Removal .......................................

3.5.3 Validation of NDE Round Robin Results ..................

3.5.4 Burst Testing of Service-Degraded Tubes ...............

3.6 INSERVICE INSPECTION SAMPLING PLAN EVALUATION .................

3.6.1 EC Sizing Error and Probability of Plugging ...........

3.6.2 Probability of Exceeding EC Plugging Limit (PEL) .....

3.6.3 Probability of Detection (POD) .........................

3.6.4 Evaluation and Comparison of

Sampling/Inspection Schemes ........................

3.6.5 Monte Carlo Simulation Analysis ........................

3.7 INTERGRANULAR STRESS CORROSION CRACK ROUND ROBIN .... o .........

3.8 COOPERATIVE RESEARCH NOT FUNDED BY PROJECT.................

4.0 CONCLUS'IONS .................................................

4.1 TUBE INTEGRITY...........................................

4.2 NDE RELIABILITY ...............................................

4.3 EVALUATION OF SAMPLING PLANS FOR ISI ..........................

4.4 STEAM GENERATOR PREPARATORY TASKS AND INITIAL INSPECTIONS .....

5.0 RECOMMENDATIONS FOR IMPROVED ISI AND TUBE PLUGGING LIMITS ..........

5.1 RECOMMENDATIONS FOR IMPROVED ISI ..........................

5.1.1 Full-Length Tube Inspection ............................

5.1.2 Screening of Data Prior to Analysis.................

5.1.3 Independent Review of Data Analysis ....................

5.1.4 Sampling Plan ..........................................

5.1.5 Application of Frequency Mixes for Noise

and Foreign Deposit Suppression ........................

3.50

3.51

3.53

3.59

3.86

3.90

3.92

3.92

3.94

3.95

3.98

3.108

3.114

4.1

4.1

4.3

4.5

4.6

.5.1

5.1

5.1

5.1

5.1

5.1

5.3

xiv
5.1.6 Development of Performance Demonstrations for

Personnel,. Equipment and Procedures ....................

5.1.7 Improved Techniques for Crack Detection ................

5.2 RECOMMENDATIONS FOR IMPROVED TUBE PLUGGING LIMITS .............

5.2.1 Through-Wall Flaws .....................................

5.2.2 Disposition of Crack-Type Flaw Indications .............

5.2.3 Flaw Characterization and Consideration of

Inspection Uncertainties ...............................

5.2.4 Methods for Flaw Evaluation ............................

5.2.5 Repair Versus Plugging .................................

5.2.6 Plugging Limit for Allowable Denting ...................

5.2.7 Flaw Growth Allowance ..................................

6.0 REFERENCES .........................................................

APPENDIX A - COMPENDIUM OF DEFECT GROWTH RATES ..........................

APPENDIX B - STEAM GENERATOR TUBE INTEGRITY PROGRAM/STEAM GENERATOR

GROUP PROJECT REPORTS AND PUBLICATIONS .....................

5.4

5.4

5.4

5.4

5.4

5.4

5.5

5.7

5.7

5.8

6.1

A.1

B.1

xv

FIGURES.-

1.1 EDM-Slots, Elliptical-Wastage, and Uniform-Thinning

Wastage Flaw Specimens.................... .............. I 1.3

1.2 Normalized Burst Pressure vs EDM Slot Length for 0.875-in.

x 0.050-in. Tubes ............................................. 1.4

1.3 Normalized Burst Pressure vs Uniform-Thinning Length for

0.875-in. x 0.050-in. Tubes ...................................... 1.4

1.4 Normalized Burst Pressure vs EDM-Slot Depth for 0.875-in.

x 0.50-in. Tubes ......... ..................................... 1.5

1.5 Normalized Burst Pressure vs Uniform-Thinning Depth for

0.875-in. x 0.050-in. Tubes ...................................... 1.5

1.6 Normalized Burst Pressure vs Elliptical-Wastage Depth for

0.875-in. x 0.050-in. Tubes ...................................... 1.6

1.7 Predicted Normalized Burst Pressure (EDM-Slot Eqn.) vs

Measured Normalized Burst Pressure for 0.875-in. x

0.050-in. Tubes .................................... ........... 1.7

1.8 Predicted Normalized Burst Pressure (Uniform-Thinning Eqn.)

vs Measured Normalized Burst Pressure for 0.875-in. x

0.050-in. Tubes .... ........................................... 1.7

1.9 Predicted Normalized Burst Pressure (Elliptical-Wastage Eqn.)

vs Measured Normalized Burst Pressure for 0.875-in. x

0.050-in. Tubes ............................................... 1.8

1.10 Comparison of Burst Pressure Equations for Elliptical

Wastage, Uniform Thinning, and EDM Slots ........................ 1.9

1.11 Normalized Burst Pressure Predicted From the ASME

IWB-3640 Axial Flaw Equation vs Actual Values for

Phase I Machined Flaws ........................................... 1.10

1.12 EC-Indicated vs Actual EDM-Slot Depth ........................... 1.12

1.13 EC-Indicated vs Actual Elliptical-Wastage Depth .................. 1.12

1.14 EC-Indicated vs Actual Uniform-Thinning Depth .................... 1.13

1.15 EC-Indicated vs Actual EDM-Slot Length.......................... 1.14

1.16 EC-Indicated vs Actual Uniform-Thinning Length ................... 1.14

2.1 Predicted Normalized Burst Pressure-(Phase I EDM-Slot Eqn.)

vs Measured Normalized Burst Pressure for IGSCC-Flawed Tubes ..... 2.2

xvi'i
2.2 Predicted Normalized Burst Pressure (Phase I UniformThinning Eqn.) vs Measured Normalized Burst Pressure for

Chemically-Produced Uniform Thinning ............................. 2.2

2.3 Comparison of Hall and Powell (1987) and Berge (1987)

Leak-Test Data With Model Predictions, Normal Operating

Condition ......................................................... 2.5

2.4 Comparison of Battelle-Columbus Leak-Test Data With Model

Predictions, Test #10, Axial Crack .............................. 2.6

2.5 EC-Indicated vs Actual SCC Depth ................................. 2.7

2.6 EC-Indicated vs Actual Elliptical-Wastage Depth (Electroetch) .... 2.8

2.7 EC-Indicated vs Actual Uniform-Thinning Depth (Electroetch) ...... 2.8

3.1 Surry Steam Generator Upon Arrival at Hanford .................... 3.5

3.2 Cut-Away of SGEF Showing Position of Steam Generator ............. 3.6

3.3 Radiological Maps - Floors 1-4 SGEF .............................. 3.8

3.4 Relative Radiation Level Along a Typical Hot Leg Tube - R22C47... 3.10

3.5 Relative Radiation level Along a Typical Cold Leg Tube - R22C47.. 3.11

3.6 Fe, Cr, Ni, Mn, and Co-60 Concentrations in the Reagent

During the Hot Leg Decontamination ............................... 3.13

3.7 Fe, Cr, Ni, Mn, and Co-60 Concentrations in the Reagent

Before Ion Exchange During the Cold Leg Decontamination .......... 3.14

3.8 Sectioned Explosive Plug in Simulated Tube Sheet ................. 3.16

3.9 Surry 2A Steam Generator Tube-Plugging Map as of 8/8/78 .......... 3.17

3.10 Schematic of Model D Plug Drilling Fixture ....................... 3.19

3.11 Location of Water and Sludge Samples ............................. 3.20

3.12 Cameras Fabricated for Use in Surry Generator .................... 3.23

3.13 Side Movement (Parallel to Lane) of Tubes Relative to the

First Support Plate, Column 39-40 ................................ 3.24

3.14 Probable ID (Primary) Side Initiated Crack Which Appears

as an Axial Crease in the Apex of Column 87 ...................... 3.25

3.15 Montage Photo of the Top of the Seventh Support Plate

at the Tube Lane, Column 9 to Column 1 ........................... 3.26

xviii
3.16 Probable Burst-Type Fracture, U-Bend RIC85; a) Penetration.-

Below Seventh TSP, b) Looking Up (Intrados Region),

c) Looking Down (Extrados Region) ................................ 3.27

3.17 Bottom Side of First Support Plate, From O°.Handhole ............. 3.29

3.18 Montage Photo Looking Up at the Second Support Plate From

Shell Wall (Hot Leg Side) to the Cold Leg Side at

Column 62-63 ................................................. 3.30

3.19 Overview and Sludge Pile Closeup of Hot Leg Tube'Sheet

Section #355 .................................... . I........ .... 3.32

3.20 Overview and Sludge Pile Closeup of Cold Leg Tube Sheet

Section #354 .................................. !. ........... .3.33

3.21 Subset of Tubes Selected for Round Robin Inspection-s....,...-,......... -.3.39

3.22 Abrasive Cut-Off Wheel ........................................... 3.54

3.23 Semi-Remote Abrasive Wheel,Cutting at Top of Seventh TSP... .... 3.54

3.24 Full-Length Tube Being Pulled Up by SGEF Bridge Crane ........... :3.56

3.25 Tube Pulling Through Tube Sheet; Hydraulic Jack in Operation ..... 3.57

3.26 Tube Sheet Section Drilling Equipment in Place-.in Channel Head ... 58

3.27 Wear and Corrosion at AVB Contact Areas on U-Bend

Specimens 416A and 416B (R31 C58). ............................... .3.61

3.28 Optical Micrographs Showing Axial IGSCC at the Apex of

U-Bend Specimen 1049 (R2 C27)..................................3.62

3.29 U-Bend Specimen 450A (R14 C52 HL) Showing Discontinuities

in the Copper-Rich OD Surface Deposit.... ........... o ......... 3.63

3.30 Typical TSP Intersection; Specimen 665 (R7 C57 HL)............... 3.65

3.31 First TSP Specimen 1041 (R4 C51 HL) Showing ID

Cracking (26.9% strain) .......... ............................... 3.66

3.32 Optical Micrograph Showing Ring Type Wastage on

Hot-Leg TTS Specimen 632 (R13 C37 HL) ............................. 3.69

3.33 Pitting/Wastage Type Degradation of a Hot Leg TTS Specimen ........ 3.70

3.34 Pitting/Wastage Type Degradation of a Hot Leg TTS Specimen....... 3.70

3.35 Optical Micrographs From a Longitudinal Section Showing OD

Circumferential Cracking of TTS Specimen 633C

(R13 C44 HL at TTS + 0.2 in.) ................................... 3.71

xix
3.36 Median Probability of Detection of DAARR and Base-Line

Inspection Teams Based on Visual Determination of Wall-Loss. ..... 3.77

3.37 Median Probability of Detection of DAARR and Base-Line

Inspection Teams Based on Metallographic Determination

of Wall-Loss ..................................................... 3.78

3.38 Probability of Detection of an AATRR Inspection Team Based

on Metallographic Determination of Wall-Loss .................... 3.78

3.39 Individual POD Performance for the DAARR and Base-Line

Teams for Each 10% Wall-Loss Increment ........................... 3.79

3.40 Typical Plot of EC Estimated Flaw Depth Versus Actual

Depth from Metallographic Analysis .............................. 3.80

3.41 Plot of Best Observed EC Estimated Flaw Depth Versus Actual

Depth from Metallographic Analysis .............................. 3.80

3.42 Comparison EC Data Reanalyzed at Various Frequencies

With Metallography ............................................... 3.83

3.43 Correlation of EC Estimated Wall-Loss (Base-Line and DAARR

Teams) With Metallography ........................................ 3.84

3.44 Correlation of EC Estimated Wall-Loss (Base-Line and DAARR

Teams) With Metallography ........................................ 3.85

3.45 U-Bend Specimen 568 (R4 C3) Showing a Defect-Called Signal

in U-Bend Area Which Turned Out to be an Interruption in

the Copper-Rich Surface Deposits ................................. 3.87

3.46 Specimen 568 (R4 C3) After Cleaning .............................. 3.88

3.47 Calculated Normalized Burst Pressure From EDM Slot Equation

vs Measured Normalized Burst Pressure for Tubes Removed

From the HLTTS Region With P/W Defects ........................... 3.91

3.48 Burst Pressure vs Median Eddy-Current Depth for Tube

Segments From the Surry Steam Generator .......................... 3.91

3.49 Tube Map 1 ....................................................... 3.99

3.50 Tube Map 3 ....................................................... 3.100

3.51 Tube Map 6 ....................................................... 3.101

3.52 Tube Map 8 .................................................. 3.102

3.53 Tube Map 13 Predicted Surry "True State" ......................... 3.103

3.54 Simulation POD Curves 1 and 2 ....................... i ......... 3.104

xx
3.55 Simulation POD Curves 4 and 5 .................................... 3.105

3.56 Sampling Plan Effectiveness Versus Systematic Sample Size for Various Tube Maps - POD Curve 1/Sizing Model 1 .......... 3.106

3.57 Sampling Plan Effectiveness Versus Systematic Sample

Size for Various Tube Maps-- POD Curve 2/Sizing Model 2 .......... 3.106

3.58 Sampling Plan Effectiveness Versus Systematic Sample Size for Tube Maps 1 and 6 - POD Curve 4/Sizing Model 1 .......... 3.107

3.59 Sampling Plan Effectiveness Versus Systematic Sample

Size for Tube Maps I and 6 - POD Curve 5/Sizing Model 2 .......... 3.107

3.60

3.61

3.62

3.63

3.64

3.65

3.66

3.67

3.68

Summary of POD Results

EC Indicated vs Actual

EC Indicated vs Actual

EC Indicated vs Actual

EC Indicated vs Actual

EC Indicated vs Actual

EC Indicated vs Actual

EC Indicated vs Actual

EC Indicated vs Actual

for

SCC

SCC

SCC

SCC

SCC

SCC

SCC

SCC

IGSCC Round Robin T

Depth for Team MD -

Depth for Team MF -

Depth for Team MH -

Depth for Team MO -

Length for Team MD

Length for Team MF

Length for Team MH

Length for Team MO

eams...

B

Bobbin ............

Bobbin ............

RPC ...............

Array Coil ........

- Bobbin ...........

- Bobbin ...........

- RPC ..............

- Array Coil .......

3.109

3.110

3.110

3.111

3.111

3.112

3.112

3.113

3.113

5.1 40% Systematic Sampling Grid Pattern for First

Inservice Inspection .............................................

5.2 40% Systematic Sampling Grid Pattern for Second

Inservice Inspection..............................................

5.3 40% Systematic Sampling Grid Pattern for Third

Inservice Inspection ....................................

5.4 Burst Parameter Prediction Curves for Axially-Oriented

Degradation in Steam Generator Tubes. L = Degradation

Length, a = Degradation Depth, R = Tube Inner Radius,

and t = Tube Wall Thickness ......................................

5.5 Burst Parameter Prediction Curves for Circumferentially-Oriented

Degradation in Steam Generator Tubes. L = Degradation

Circumferential Length, a = Degradation Depth, R = Tube Mean

Radius, and t = Tube Wall Thickness ..............................

5.2

5.2

5.2

5.6

5.7

xxi
A.1 Average Corrosion Pene~trations of Alloy 600 in Deaerated

SodiumPhosphate Solutions a1 615 0 F, as ;a Function of the

Na/P04 Molar Ratio............................................ A.2

A.2 Temperature Dependence of theCorrosion Rate of Alloy 600

Tubing Exposed to Deaerated Phosphate Solutions .................. A.3

A.3 Pit Depthyvs Exposure Time in Alloy 600 Tubes in Model

and Pot Boiler Tests ........ ; ......... .......................... A.5

A.4 Histogram ShowingStrain MeasurementsFrom the PROFILE-360

Inspection.................................................... A.9

A.5 Temperature Dependence of SCC Initiation Strain for CERT

Tests at e =W2xlO- 1O sec- 1 . ............... .. .. ... .. .. ... .. .. ... . A.10

A.6 Temperature Dependence of IGA Growth Rate ....................... A.12

A.7 Temperature Dependence of SCC Growth Rate ........................ A.13

A.8, Influence of-Stress on IGA as a Function of Temperature .......... A.14

A.9 Stress.Corrosion Tests in,Deaerated Sodium Hydroxide at,

350!Cop Fracture Mechanics Type Specimens: Comparison

of Alloy 600 and 690 Behavior ..................................... A.15

A.10 Summary.of Corrosion- Fatigue Crack Growth Results for

Inconel 600 ...................................................... A.17

A.11 Depth-of IGA Penetrations on Mill-Annealed Alloy 600 Tubing

(Heat NX3335) as a Function of Exposure Time ..................... A.18

xxii
TABLES

1.1 Range of egradationDimensions .............................. 1.2

1.2 Phase I Single-Frequency EC Depth,-Sizing Models .................. 1.11

1.3 Phase I Single-Frequency EC Length-Sizing Models ................. 1.13

2.1 Predictive FailureEquations.......,................................... 2.3

3.1 Results From ICP Analysis of Sludge Samples From,

Tube Sheet Specimen #355 (HL) and #354 (CL) .................... 3.34

3.2 Results From X-Ray Diffraction of Sludge Samples From

Tube Sheet Specimen #354 (HL) and #355 (CL) ...................... 3.34

3.3 Locations of Wall-Loss Indications From the

Base-Line Inspections ............................................ 3.36

3.4 Numbers of Tubes in the Round Robin Sample

Listed by Size and Agreement Substrata ............................ 3.38

3.5 Numbers and Types of Indications Reported by

Acquisition and Analysis Round Robin Teams ....................... 3.41

3.6 Location of Indications Reported by Acquisition and

Analysis Round Robin Teams ................................. 3.43

3.7 Numbers and Types of Indications Reported by

Advanced/Alternate Technique Teams ............................... 3.44

3.8 Location of Indications Reported by

Advanced/Alternate Technique Teams ........................................ 3.45

3.9 Statistical Summary ofProfilometry Data .......................... 3.46

3.10 Numbers and Types of Indications Reported by

Analysis Round Robin I Teams ...................................... 3.47

3.11 Location of Indications Reported by

Analysis Round Robin I Teams ...................................... 3.48

3.12 Numbers and Types of Indications Reported by Analysis Round Robin II Teams...................................... 3.49

3.13 Location of Indications Reported by

Analysis Round Robin II Teams .................................... 3.50

3.14 Numbers of Specimens per Strata and Major

Location Within the Generator ............................... 3.52

xxiii
3.15 Summary of Metallography Result.s Regarding OD

Corrosion of Dented TSP Intersections ............................ 3.66

3.16 Summary of ID Cracking at Dented TSP Intersections

With Various Strain Levels........ ............................ 3.67

3.17 Summary of OD Visual Examinations of TTS Specimens

From the Sludge Pile Region of the Generator...................... 3.68

3.18 Summary'Classification of Individual Team NDE

Inspection Results by Region of the'Generator .................... 3.73

3.19 Sizing Regression SummaryStatistics ......................... 3.81

3.20 Surry Tube Burst Data ......... ............................... 3.89

3.21 Estimates of PEL for X = 75% and T= 40% ......................... 3.93

3.22 Plugging Limit T Required for PEL = 0.9When X = 75% ............. 3.93

3.23 Estimated PEL Values for Three Plugging Limits .................... 3.94

3.24 Values of p Computed for a Range PEL and POD(deg)

Values Assuming That POD = 0.9 for Defectives ................... 3.97

3.25 EC Sizing Models Used in Simulations.............................. 3.105

A.1 Maximum Percent Wall-Loss- for Wastage Degraded Surry Tubes ....... A.2

A.2 Number of Tubes Plugged for Wastage Degradation .................. A.3

A.3 Estimated Growth Rates for Surry Generator Wastage

Degradation (mils/yr) ............ .. ., ............................ A.4

A.4 Laboratory Pit Growth Rates (mils/yr) ......................... A.6

A.5 Maximum Percent Wall-Loss for Hot-Leg Pitting and

Pitting/Wastage Degradation of Surry Tubes ....................... A.6

A.6 Estimated Pitting and Pitting/Wastage Growth Rates from

Surry Tubes .................................................... A.6

A.7 Estimated Upper Bound Growth Rates for Cracking Mechanisms

(mils/yr) ................... ..................................... A.14

A.8 Upper Bound Growth Rates (600'F) for Principal Steam

Generator Tube Damage Mechanisms (mils/yr) ...................... A.16

xxiv
1.0 PHASE I SUMMARY

The objective of the Steam Generator Tube Integrity Program (SGTIP) was

to experimentally determine the margin-to-failure and inspection reliability

of degraded steam generator tubes. The program was initiated in 1976 by the'

NRC in response to an inquiry regarding the existence of NRC data on the

integrity of service-degraded steam generator tubes. The initial program

plan called for a three phase effort. The first phase involved burst and

collapse testing and eddy current examination of typical steam generator tubing

with machined flaws. The second phase of the SGTIP continued the integrity

testing and NDE work of Phase I, but tubes were degraded by chemical means

rather than machining methods. The third phase of the SGTIP was to validate

the results from Phases I and II by testing tubes with service-induced flaws.

This section summarizes results of Phase I of the SGTIP. A detailed

presentation of the experimental data and interpretation may be found in the

Phase I Report by Alzheimer et al. (1979). During Phase I, segments of

Inconel 600 steam generator tubing manufactured in accordance with 1977

practice were artificially degraded, then burst and collapse tested to

determine the remaining margin-to-failure as a function of degradation type

and size. The four common types of steam generator tube sizes used in the

United States were tested. All specimens were tested under simulated steam

generator operating conditions. In addition:, single-frequency eddy-current

(EC) examinations were performed on each degraded tube segment to assess .the

capability of EC techniques to detect and size machined wall-loss degradation.

1.1 TUBE INTEGRITY TESTS

Three forms of machined degradation were employed in the Phase I work.

These included uniform-thinning, elliptical-wastage (localized wall thinning)

and an electrodischarge-machined (EDM) slot representation of crack-like

degradation. Figure 1.1 gives a schematic of each degradation type and

Table 1.1 lists the range of dimensions used for each degradation type.

After machining each degradation type in each specimen, the dimensions

of the degraded zone were determined by plastic replication. Specimens were

tested at Pressurized Water Reactor (PWR) steam generator operating

temperatures under controlled loading conditions. In all, over 500 specimens

were tested.

Burst-test results on 0.875-in. diameter x 0.050-in. wall-thickness tubing

with each type of degradation are given in Figures 1.2 to 1.6. These data

typify the results for all tube geometries tested. The burst data were

nondimensionalized by dividing the burst pressure of a degraded tube by the

burst pressure of a nondegraded tube to minimize the effect of material

property variations. Figures 1.2 and 1.3 show the effect of increasing

degradation length on normalized burst pressure for different depths of

EDM-slot and uniform thinning, respectively. The effect of variable

elliptical-wastage length was not examined in this study, so no plot is

presented. The normalized burst pressure was observed to decrease rapidly as

1.1
TABLE 1.1. Range of Degradation Dimensions

Depth, P0Length,

in.

End

Flaw Type Radius, in.

Wrap

Anale. 0

EDM Slots

Elliptical Wastage

Uniform Thinning

25-30

55-60

80-90

25-30

55-60

80-90

25-30

55-60

70-80

0.25

0.5

1.5

P1.5

0.188

0.375

0.75

1.5

0.01-0.02

NA

0.063

NA

0

45

135.

NA

NA - Not Applicable.

degradation length increased. This effect tended to saturate for degradation

lengths greater than or equal to about one tube diameter. Figures 1.4 to 1.6

give plots of normalized burst pressure as a function of degradation depth

for different degradation lengths. It is evident from these plots that the

burst pressure decreased with increasing depth of degradation for fixed

degradation length. The elliptical-wastage data presented in Figure 1.6 was

partitioned with respect to wrap angle (i.e., circumferential extent of the

degraded zone) to determine if this variable had an affect on the resulting

burst pressure. No significant effect was observed for the range of wrap

angles investigated.

Empirical equations were developed from the test data by least squares

curve-fitting techniques. Theicurve fit equations developed were:

For EDM Slots

AP/APo = 1 - (h/t) + (h/t)exp{-0.373 L/IVRt}

For Uniform Thinning

AP/APo = (1 - h/t)l - exp{-0.13 LIVR(t-h)}

(1.1)

(1.2)

(1.3)

For Elliptical Wastage

AP/APo : (0 - h/t)0. 6 0 4

1.2
IImzzz~ I - -.

/ N 'I

\ / - -

57: 0

(a) EDM SLOT (b) ELLIPTICAL WASTAGE

- - - N

/

I (p I

(c) UNIFORM THINNING

FIGURE 1.1. EDM-Slots, Elliptical-Wastage, and Uniform-Thinning

Wastage Flaw Specimens

1.3
1.0

U)

=3 0.8-

/)

a)

IL. 0.6

, 0.4- N

o 0.2-

Z

0.0

13

ci3

00

0 0

A

0

0

Slot Depth

0 13 25-30%

A 55 - 60%

0 80-90%

4

0.0

i i

0.5 1.0

EDM Slot Length, in.

1.5 2.0

FIGURE 1.2. Normalized Burst Pressure vs

EDM Slot Length for 0.875-in.

x 0.050-in. Tubes

1.,U

co

:3

a)

E)

0)

a0.8-

0.6-

0.4-

0.2-

B

A [

A 130 2

8

UT Depth 0

13 25-30%

A 55-60%

o 70-75%

A A

0.0 0.5 1.0 1.5

Uniform Thinning Length, in.

2.0

FIGURE 1.3. Normalized Burst Pressure vs

Uniform-Thinning Length for

0.875-in. x 0.050-in. Tubes

1.4
1.0

CL.

(0

(D

L4.

0~

.4-

0.8-

0.6-

0.4-

0.2-

13

13

"al

Slot Length A

[3 0.25 in.

a 0.5 In. 0

0 1.5 in.

0.0 I

0

2

20 40

EDM Slot

60

Depth, %

80 100

FIGURE 1.4. 'Normalized Burst Pressure vs

EDM-Slot Depth for

0.875-in. xk0.50-in,. Tubes

1.0

()

U) 0.8-

u)

a)

a.. .0-

: 0.6-

"O

N

0.4

0

z

(9 13El

A 0 ..... A ,A

0 0:

UT Length 0A

[3 3/16 in.

A 3/8 in.

O 3/4 in.

* 1 1/2 in.

n 1 .4 - I - I I I

20 30 40 50 60

Uniform Thinning Depth, %

7

70 80

FIGURE 1.5. Normalized Burst Pressure vs

Uniform-Thinning Depth for

0.875-in. x 0.050-in. Tubes

1.5
1.0

0.8

UI)

CL

S0.6-

~~0

( 0.4

Wrap Angle 00

OZ 0.2 A 450

0 135°

0.0 , , " .

0 20 40 60 80 100

Elliptical Wastage Depth, %

FIGURE-1.6. Normalized Burst Pressure vs

Elliptical-Wastage Depth for 0.875-in. x 0.050-in. Tubes

where AP/APo = Ratio of defected to undefected burst pressures

h = Defect depth

t =Wall thickness

R = Inner radius of tube

L = Defect length

These equations are applicable to all of the different sizes of steam

generator tubingtested. In addition, three different strength levels of a

single tubing size were tested and the equations are valid for tubing with

yield strengths ranging from 40 ksi to 48 ksi. Figures 1.7 to 1.9 give plots of the predicted normalized burst pressure versus the measured normalized burst

pressure for the 0.875-in. diameter x 0.050-in. wall-thickness tubing degraded

with EDM slots, uniform thinning and elliptical wastage, respectively. The

diagonal line in the plots corresponds to the line of perfect agreement

between predicted and measured values. Note the predicted burst pressures

for uniform-thinning and elliptical-wastage degradation agreed quite closely

with measured values. In contrast, the EDM slot data displayed more scatter.

This was due to greater variation in the, uniformity of the EDM slot

degradation. Some of the EDM'slot specimens had fairly straight walls and

flat bottoms; whereas, other specimens had slot widths and depths that were not

uniform along the length.

1.6
1.0

a) o .8

a._

im 0.6

-0

CD)

"CO

.E 0.4

0

z

& 0.2

I..

0.0

0.0 0.2 0.4 0.6 0.8

Measured Normalized Burst Pressure

1.0

FIGURE 1.7.

1.0 -

O...

U) 0)0.8-

:3

CD 0.6'

(Da)

N

O 0.4

z

0

S0.2-

L.. a-o 0.0

0.0

FIGURE 1.8.

Predicted Normalized Burst Pressure (EDM-Slot

Eqn.) vs Measured Normalized Burst Pressure for

0.875-in. x 0.050-in. Tubes

* I * . I

0.2 0.4 0.6 0.8

Measured Normalized Burst Pressure

1.0

Predicted Normalized Burst Pressure (UniformThinning Eqn.) vs Measured Normalized Burst

Pressure for 0.875-in. x 0.050-in. Tubes

1.7
L..w

CO. V)

0.86

E 0.4-

0

Z

0.2-

(D

0.0-

.00 .6 0.8 1.0

Measured Normalized Burst Pressure

FIGURE 1'.9. Predicted Norma-lized Burst Pressure (EllipticalWastage Eqn.)'vs Measured Normalized Burst

Pressure for 0.875-in. x O.050-in. Tubes

it is instructive to compare the burst pressure equations for the above

flaw types versus• flaw length for three through-wall depths, as shown in

Figure 1.10. Note the similarity of the EDM slot and uniform-thinning

equations over the full range of defect dimensions. Conversely, the

relationship for elliptical-wastage flaws consistently gives higher indicated

burst pressures for the same length and depth of flaw compared to an EDM slot

or a uniformly-thinned specimen.

A plobt of the measured normalized burst strengths for all three flaw types

versus values predicted from the tab-les presented in Section XI, Subsection

IWB-3640 of the American Society of Mechanical Engineers (ASME) Code (1986)

it shown in Figure 1.11. The replicated flaw depths and lengths were

substituted into the ASME equation to compute the predicted burst strengths.

It is evident from the results that the ASME equation gives conservative

results in most cases, especially when the actual normalized burst pressure

is below 0.5. The ASME equation was developed for analysis of axial flaws in

degraded stainless steel piping. Hydraulic failure of stainless steel piping

is controlled by net section yielding of the unflawed ligament. Thus, it

seems reasonable to postulate that Incone] 600 steam generator tubes fail by

the same mechanism.

1.8
(3)

o0.60-

4-,

N '0.40

0.20 Elliptical Watage

.- .Uniform . Thinning-.

-. EDM Slots

0.001l

- 0.0 0.2 "0.4 M06 0.8 1.0 1.2 1.,

Defect Length, in.,

FIGURE1.10. Comparis'on of Burst Pressure:Equations for<.E!.liptical

Wastage, Uniform Thinning, and EDM Slots

4 1.6 1.8
1.0

Flaw Type

C o-" EDM Slots

-• 0.8 * Unit. Thinning * El

o0 Ellip. Wastage 0 o

CO '-El

0.6.

SNonconservative

a)

• • 0

E • 0.4- 0 b 0,d 0 B Conservative Z 3: G= • 0 

W 0.2

0.0 A

0.0 0.2 0.4 0.6 0.8 1.0

Measured Normalized Burst Strength

FIGURE 1.11. Normalized Burst Pressure Predicted from the

ASME IWB-3640 Axial Flaw Equation vs Actual

Values for Phase I Machined Flaws

1.2 SINGLE-FREQUENCY EDDY-CURRENT RESULTS

After developing the above constitutive equations between defect size/

geometry and remaining tube strength, the next step was to relate eddy-current

inservice inspection results to the tube failure strength. It is desirable

to know how accurately EC measures the actual flaw size. Knowledge of EC

error would result in appropriate use of the above tube integrity equations.

The Phase I nondestructive testing and evaluation effort focused on

single-frequency (400 kHz) eddy-current examination of each defect type. An

EM-3300 test instrument, a Gould Brush 220 strip-chart recorder and a HewlettPackard magnetic tape recorder were used to acquire the EC inspection data.

Several differential bobbin-coil probes were used for the inspections,

depending on tubing diameter. Fill factors ranged from 0.78 to 0.87. Each

probe was calibrated using an appropriate ASME flat-bottom hole standard for

each tube diameter. During inspection the specimens were positioned

horizontally and examined by translating the probe through the tube until a

scan of the degraded zone was completed. A recording of the instrument

response was made as the probe was pulled back through the tube at a constant

rate. Nonsymmetrical degradation (i.e., degradation not uniformly distributed

around the circumference of the tube such as elliptical wastage and EDM slots)

was examined in three orientations. The specimen was initially placed with the

degraded area oriented at 00 (zenith); then the tube was successively rotated

and tested with the degraded zone in the 900 (horizontal) and 1800 (downward)

1.10
positions, respectively. The results from-the three test records were averaged

to minimize the effect of probe-to-degradation spacing and variations caused

by probe wobble and misalignment.

The EC estimates of EDM-slot, elliptical-wastage and uniform-thinning

depth for all tubing'geometries versus actual degradation depth are plotted in

Figures 1.12 to 1.14., It is evident from these figures that the accuracy of

degradation depth sizing depended on the volume of metal removed by the

degradation mechanism (i.e., degradation volume). For small-volume

degradation such as EDM slots, and elliptical wastage, EC overestimated the

depth when the degradation was shallow. As the-actual depth increased, the

EC results reverted from an overestimate of depth to an underestimate. The

degree of undersizing continued to increase with increasing degradation depth.

For large volume degradation, such as uniform thinning, EC consistently

overestimated flaw depth.

To quantify these trends, a linear regression, Y = A(X) + B, was curve

fit to the data,, where Y is the EC estimated degradation depth and X is the

actual depth. Table 1.2 gives the results from these calculations. The column

denoted "Number" in the table refers to the number of data points in the curve

fit, and the column denoted "R" 'gives the-correlation coefficient. It is

evident from these results that the smallest volume degradation type, EDM

slots, yielded the highest intercept and'lowest slope. The intermediate

volume degradation, elliptical wastage, gave the smallest intercept and an

intermediate slope. The large volume degradation type, uniform thinning,

produced an intermediate intercept and the highest slope. A plot of these

equations shows that oversizing was greatest for the EDM-slot flaw for

degradation less than 20% through-wall, with the elliptical-wastage flaw

oversized the least for flaws in this range. Uniform thinning was on the

average oversized at all depths, with the magnitude decreasing from 16% at

shallow flaw depths to about 6% for the deepest thinning. Undersizing of EDM

slots and elliptical wastage began at about 60% through-wall penetration and

increased with increasing flaw depth, with EDM slots being undersized more

than elliptical wastage of equal depth. The EC results on machined flaw types

clearly illustrate the difficulty of accurately measuring flaw depth for EDM

slot and elliptical wastage flaw types. This is an important result since

as flaw depth increased, the EC estimated depth was increasingly

nonconservative.

TABLE 1.2. Phase I Single-Frequency EC Depth-Sizing Models

Type Intercept Slope Number R

EDM Depth 26.7 0.52 119 0.65

EW Depth 11.5 0.77 114 0.88

UT Depth 17.4 0.87 62 0.95

1.11
100

61 80

Q 60

0

(D 40

C.)

C

0 20

w

0

0 20 40 60 80

EDM Slot Depth, %

100

FIGURE 1.12. EC-Indicated vs Actual

EDM-Slot Depth

100

0

-C

0~

w

80

60

40

20

0

0 20 40 60 80

Elliptical Wastage Depth, %

100

FIGURE 1.13. EC-Indicated vs Actual

Elliptical-Wastage Depth

1.12.
100

o 60

0O 40 U

C

020

0 20 40 60 80 100

Uniform Thinning Depth, %

FIGURE 1.14. EC-Indicated vs Actual Uniform-Thinning Depth

The accuracy of degradation length sizing was also investigated from the

Phase I EC data as shown in Figures 1.15 and 1.16 for EDM slots and uniform

thinning, respectively. The results indicate the average EC estimated flaw

length was quite close to the actual value. This is reflected in the linear

regression results shown in Table 1.3. Note the equations for both flaw types have very similar intercept and slope numbers.

TABLE 1.3. Phase I Single-Frequency EC Length-Sizing Models

Type Intercept Slope Number R

EDM Length -0.04 1.03 169 0.96

UT Length -0.02 1.02 42 1.00

It was mentioned above that tube burst strength depended strongly on flaw

length. Based on the above equations one would conclude that length sizing

of both small and large volume defects is very accurate, since for degradation

greater than 0.6-in. long the difference between calculated and actual values

is less than *3%. However, the EC data on EDM slots display considerably

more scatter than for uniform-thinning degradation as depicted in Figure 1.15.

1.13
2.0

C.

.c 1.5

'2

-J0CD

._o

0.5.

0.0 4"

0.0

0.5 1.0 1.5

EDM Slot Length, in.:,

2.0

FIGURE 1.15. EC-Indicated vs Actual

EDM-Slot Length

2.0

.E 1.5

_j

-JI--

D 1.0

"0

U) 0.0 ~0

0.0

0.0 0.5 1.0 1.5

Uniform Thinning Length, in.

FIGURE 1.16. EC-Indicated vs Actual

Uniform-Thinning Length

2.0

1.14
For example, slots 0.5-in. long had EC estimated lengths ranging from 0.228 in.

to 1.545 in., and for 1.5-in. slots the range was between 0.880 in. and 1.760

in. Thus, it may be concluded that length sizing of small- and large-volume

machined degradation was fairly accurate, but measurement precision was

acceptable only for large volume degradation.

1.15

2.0 PHASE II SUMMARY

This section summarizes the results of Phase II of the SGTIP. The

objectives of Phase II were to perform burst, collapse, leak-rate and EC

testing of representative steam generator tubing with chemically-induced

degradation to validate and refine the empirical relations developed during

Phase I. It was considered necessary to test the Phase I constitutive

equations against specimens containing flaws more closely simulating actual

service-induced degradation. Of particular importance was the effect of the

methods used to introduce the tube degradation on the reliability of EC to

characterize those flaws. Phase II of the SGTIP utilized a considerably

reduced number of specimens compared with Phase I. The flaw geometries and

sizes represented a subset of the Phase I specimen matrix. This section

summarizes the tube integrity, leak-rate and nondestructive evaluation results

for this portion of the program. A more detailed discussion may be found in

a report by Kurtz et al. (1988).

2.1 TUBE INTEGRITY TESTS

The tube integrity specimen matrix for Phase II consisted of 86

chemically-degraded tubes from material heats B, E and F. The tubing from

these heats were all 0.875-in. diameter x 0.050-in, wall thickness. It was

felt that flaws produced by chemical methods more closely approximate serviceinduced flaws with respect to variable shape, size, depth and orientation.

Three flaw geometries were investigated: stress corrosion cracking (SCC),

uniform thinning and elliptical wastage. The flaws were created by exposing

segments of tubing to various aggressive environments. Flaw shape, location

and orientation were controlled by masking the tubing with Teflon )(a) tape.

Results from burst testing of SCC and uniform-thinning degraded specimens

are plotted in Figures 2.1 and 2.2, respectively. The measured normalized

burst pressures are plotted against calculated values from the appropriate

Phase I empirical relationship. These data are typical of Phase II results.

As shown in Figure 2.1, the EDM slot constitutive equation, in almost all

instances, yielded conservative estimates of remaining tube strength when

bounding crack dimensions were used in the evaluation. Figure 2.2 gives

burst-test results for uniform-thinning specimens flawed by two different

procedures. No significant difference due to the defecting mechanism is

evident from the burst-test results. Further, the Phase I uniform-thinning

burst equation closely predicted the burst pressure of chemically-flawed tubes

with a slight nonconservative trend. This result was typical for

elliptical-wastage specimens as well.

The data from Phase II Were used to update the Phase I equations. The

functional form of the equations was not changed. Wherever appropriate, the

constants in the predictive relationships were recalculated by least squares

regression analysis using the Phase II data alone and in conjunction with the

(a) Teflon is a. registered trademark of E. I. du Pont De Nemours and Co.,

Inc.,. Wilmington, Delaware.

2.1
IA

a)

CO)

(1)

a)

"0.8

U)

m0.6

a)

.N

E 0.4

0

z

0.2

a)

a_.

0.

0.0 0.2 0.4 0.6 0.8

Measured Normalized Burst Pressure

1.0

FIGURE 2.1. Predicted Normalized Burst Pressure

(Phase I EDM-Slot Eqn.) vs

Measured Normalized Burst Pressure

for IGSCC-Flawed Tubes

1.0

:D

ILa)

a)

CO 0.8

E 0.4

a)

I.-

0.0

0.0 0.2 0.4 0.6 0.8

Measured Normalized Burst Pressure

1.0

FIGURE 2.2. Predicted Normalized Burst Pressure

(Phase I Uniform-Thinning Eqn.) vs

Measured Normalized Burst Pressure for

Chemically-Produced Uniform Thinning

2.2
Phase I data. Since collaps~e data for the elliptical-wastage flaw type were

very sparse and nonexistent for SCC specimens, the Phase I relationships were

not changed. Table 2.1 lists the functional form of the six empirical

relationships obtained along with the constant C.

TABLE 2.1. Predictive Failure Equations

Values of C

EDM Slot - Burst Phase I Phase II Phase I & II

AP/APo = 1 - h/t +h/t exp{CL/ARt} -0.373

Elliptical Wastage - Burst

AP/APo = (1 - h/t)C 0.604 0.700 0.626

Uniform Thinning - Burst

AP/APo = (1 - h/t)l - exp{CL/VR (t - h)} -0.130 -0.200 -0.142

EDM Slot - Collapse

AP/APo = 1 - exp{CL/ýRt} -2.49

Elliptical Wastage - Collapse

AP/APo = (1 - h/t)C 0.396

Uniform Thinning - Collapse

AP/APO = 1 - h/t + h/t exp{CL/A/R (t - h)} -0.066 -0.118 -0.079

The definitions of the variables in the above equations is the same as

given in Section 1.1.

2.2 LEAK-RATE TESTS

This section briefly describes a survey of data from leak-rate tests on

tubes with laboratory-produced SCC. The objective of this work was to

determine if a crack will leak at a rate that is consistent with the length

of crack and the fluid pressure differential across the tube wall. The

accuracy of leak-rate predictions is an important consideration in evaluation

of leak-before-break analyses and in the determination of leak detection

limits as they relate to allowable leak rates given in plant technical

specifications.

Three sets of experimental data were utilized in the evaluation. Data

from Powell and Hall (1987) and Berge (1987), and data from Battelle-Columbus

Laboratories produced under subcontract to the SGTIP. A simple predictive

2.3
model was developed to aid in' the interpretation and comparison of the data

from the various laboratories. This''model Was adapted from one developed for

leakage from axial cracks in reactor Pressure ve.ssels described by Simonen

et al. (1986). The model used fracture mechanics concepts to predict the

crack opening area as a function of crack length and pressure differential

across the tube wall. Given this crack opening area, well-known equations

from the fluid mechanics literature for flow through-an elongated orifice

were then used to predict leak rates.

Figure 2.3 shows a representative plot of the leak-rate data at normal

reactor operating conditions as a function of axial crack length. Curves

giving predicted leak rates are also plotted for comparison with the

experimental data points. Powell and Hall (1987) reported crack lengths as

bounding values; in Figure 2.3 these are indicated as a line. In examination

of the Powell and Hall (1987) data, it was apparent that a large number of

tests had a reported leak rate of exactly 0.1 gal/min. This was interpreted

as an indication on the sensitivity of the flow rate measurements. It should

be noted that for purposes of clarity in displaying the data in Figure 2.3,

many of the data points were slightly offset from the 0.1 gal/min value to

prevent superposition of data. In addition, Berge's (1987) data were not

reported in a tabular format suitable for plotting, so the bounding correlation

curves were plotted in Figure 2.3.

The data show a large degree of uncertainty in expected leak rates. This

scatter may be substantially reduced by plotting the data as a function of

the average crack length. Even with this refinement, some of the data points

would still remain as much as a factor of ten from the expected trend line.

Some of the reasons for the unpredictability of leak rates may be: 1) small

crack openings that could be plugged by impurities in the water; 2) residual

stresses from fabrication and precracking of the tube, which could cause either

crack opening or closure; 3) crack roughness, which can only be approximately

estimated and is subject to considerable.variation; and 4) the lengths were

reported with a wide range of uncertainty, and no doubt differed from the

ideal through-wall cracks used in the model.

A representative plot of leak-rate data obtained from experiments

conducted at Battelle-Columbus Laboratories (BCL) is shown in Figure 2.4.

The results of the BCL tests are presented in an alternate format, since

complete traces of pressure, temperature and flow rates for a full range of

pressure differentials could be compared with predictions. In Figure 2.4,

the measured flow rate as a function of time is shown. The length of axial

crack in the specimen was determined as between 0.5 in. and 1.0 in. The

end-of-test flow rate reached a maximum level of about 1.6 gal/min,

corresponding to a pressure in the range of 2200 psi (simulating a steam line

break condition). For the lower bound crack length of 0.5 in., the predicted

flow rates were somewhat less than those measured in the test. However, the

differences between prediction and measurement were well within the variation

expected on the basis of the uncertainty in crack length. For example, when

the upper bound of 1.0 in. was used for the crack length, the model over

predicted the flow rate by a factor of about 60. The measurement and

prediction would be in very good agreement for an assumed crack length of

0.6 in.

2.4
10 Plastic Analysis

1

Predicted

ca

to

Elastic Analysis

Outside

4 Average

-'Inside

0e le French Data

-.--.--- -,,Range of Crack Length

as Measured from ID to OD

0.10

/

-0.01 I0.001 I I I I I

0 0.2 0.4 0.6 0.8 1.0

Crack Length, inch 38808040.2

FIGURE 2.3. Comparison of Hall and Powell (1987)

and Berge (1987) Leak-Test Data With Model

Predictions, Normal Operating Condition

2.5
E

0

Ua)

1.8

1.6

1.4

1.2

1.0

0.8

0.6

- 180

0

- 160

- 140

- 120 C

Em

100 •

0

80 L,

U

- 60 -5

- 40

20

_- 0

26.0

N) ýh

0.4 -

0.2 -

0o.

2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 18.0 20.0 22.0 24.0

Time (min)

38808040.1

FIGURE 2.4. Comparison of Battelle-Columbus Leak-Test Data

With Model Predictions, Test #10, Axial Crack
The main conclusion of this evaluation was that leak rates-measured during

tests can be highly variable. These leak rates can be strongly influenced by "random" variables'that are not addressed in the predictive models that have

formed the basis of leak-before-break evaluations. Measured leak rates are

sometimes up to a factor of ten less than the predicted rates. This suggests

that a substantial level of conservatism should be applied to the ,predictions

of leakage that are used for evaluations of leak-before-break. Such

conservatism 'would also appear appropriate for the calculations used to

establish leak detection limits for detection systems.

2.3 SINGLE-FREQUENCY EDDY-CURRENT RESULTS

As i:n the Phase I work, an i.ntegral ,part of the Phase II effort involved

single-frequency EC inspection and evaluation ,of the chemically-degraded tube

specimens. Thi-s section discusses 'the EC inspection results obtained 'by PNL

NDE specialists on the SCC- and wastage-degraded tubes. Section 3.7 discusses

a mini-round robin conducted on a subset of the.Phase II 'SCC-degraded tubes

conducted •dur-Ing 1987 to supplement results from round robins performed on

the Surry generator.

Si'ngle-frequency EC :measurements were made with Zetec MIZ-'7

single-frequency test equi-pment. The system consisted of an Automation

Industries EM-3300 2-channel !EC tester, a Goulid Brush 220 2-channel :stri~p-chart

recorder, a Teac 4-2300SX tape recorder and a 2-rcoil differential -wound lbobbi~n

probe of 0.750-in. diameter. -Results for these tnspections are presented in

Figures 2.5 to 2.7 for SCC, elli-ptical-wastage 'and uniform-thinning specimens,

respectively. The results roughly correspond to thýe 'Phase I data,, in that

small-volume fl'aws were undersized while large-volume flaws were oversized.

The trend seen in the Phase I data concerning oversizing of shallow,

small-volume flaws and undersizing of deep, small-volume'fl~aws was not observed

in the Phase II 'results. In addition, the small'est volume flaws, SCC,'were

depth sized with the least precision.

100

60

60, 0 0

00

0 20 40 60 .80 00

Actual CrackDepth, %

FIGURE 2.5. EC-Indicated vs-Actual SCC Depth

2.7
100.

= L

I C

I-a-* 80 LL. a

4-LU ~Z60 -

• 40

ir•

.-)

20

LUj

0

0 20

ACTUA

FIGURE 2.6.

100

UNIFORM

(ELECTRO

IL 80

z" 60

LU

LU

_z_0

0 40 z-

-~ 40

20

0

20 0- 

ACTUA

FIGURE 2.7.

40 60 80 100

L FLAW DEPTH (% WALL THICKNESS)

EC-Indicated vs Actual

Elliptical-Wastage Depth

(Electroetch)

40 60 80 100

L FLAW DEPTH 1% WALL THICKNESS)

EC-Indicated vs Actual

Uniform-Thinning Depth

(Electroetch)

2.8
3.0 PHASE III SUMMARY

Phase III of the original SGTIP was to have been a verification of Phase I

and II research results utilizing service-degraded tubing. The objectives

were to validate the flaw geometry assumptions, the empirical models of

remaining tube integrity and the EC test results developed during Phases I

and II. The Phase III research effort .as originally planned could not be

conducted, however, since suitable samples of service-degraded tubing were

generally unavailable. The small amount of removed-from-service tubing that

could be located was in short sections and usually not flawed. Those

specimens that were flawed often exhibited damage due to removal. Many others

had been metallographically sectioned for degradation mechanism

determinations.

Concurrent with the research in Phase II of the SGTIP, new forms of

service degradation were gaining public attention, such as intergranular attack

in the tube sheet crevice, and primary-side initiated cracking in roll

transitions, in conjunction with denting and in the inner row U-bends. All

of these flaws were proving difficult to detect and characterize with the NDE

techniques in use at the time. Also, at this time steam generator degradation

had progressed to the point at several units that the issue of replacement

began to be addressed. The potential for an actual service-degraded unit to

be used as a test bed was recognized by the NRC because of the unique

opportunities represented by the generator and because of lower estimated

cost compared to removal of tube samples from operational steam generators.

As a result, the retired-from-service Surry 2A Steam Generator was acquired

to conduct a research project to validate the tube integrity equations

developed in Phases I and II, and to generate an extensive data base on the

reliability of conventional and near-term field practice EC inspection

techniques to detect and size tube degradation. Phase III was organized as

the Steam Generator Group Project and included participation by the Electric

Power Research Institute and consortia from France, Italy, and Japan. This

section summarizes the results of the more than six years of research

conducted under the Steam Generator Group Project (SGGP).

3.1 SGGP OBJECTIVES AND ORGANIZATION

The primary objectives of the SGGP were to utilize a service-degraded

unit to 1) evaluate the types of degradation that had actually occurred,

2) establish the effect of this degradation on primary system integrity to

validate the Phase I and Phase II empirical models, 3) determine the

reliability of NDE inspection methods to detect and characterize the tube

degradation, 4) evaluate the reliability of near-term advanced/alternate NDE

technology, and 5) use the generator for demonstrating repair techniques.

In addition to the above objectives, the unique opportunity to study a

retired steam generator led to development of additional objectives. They

included 1) an evaluation of the usefulness of secondary-side visual inspection

for characterizing tube and support structure condition, 2) an evaluation of

sludge-pile chemistry and analysis of tube surface deposits, 3) a large-scale

unplugging of tubes by plug drilling, 4) characterization of tube sheet crevice

3.1
corrosion products by removal and~metallurgical evaluation of tube sheet

sections, and 5) an evaluation of primary-side chemical decontamination

techniques.

The program was divided into four major elements. The first element

consisted of the activities essential to preparing the steam generator for

the research program., These efforts included providing a facility, placing

the generator into the facility and making the generator ready for research.

The latter included radiological mapping and health physics activities aimed

at keeping exposures as low as reasonably achievable (ALARA), decontamination

of the channel head, removal of "inservice placed plugs and providing access

to the secondary side through the generator shell. Thesecond element involved

the data gathering phase. This included secondary-side characterization and

the acquisition of NDE data. Following these activities, the third element

consisted of a massive effort to remove segments of tubing from the generator

for validation of the NDE data and to acquire specimens for mechanical

integrity testing. The last program element consisted of visual and

destructive metallographic analysis:of the removed specimens and statistical

evaluation of the data to derive improved methods for steam generator inservice

inspection and maintenance of tube integrity.

Because examination of a retired-from-service steam generator represented

a unique opportunity, the NRC sought participation by the domestic and foreign

nuclear power industry and governments. The SGGP was formed with

participation by the Electric Power Research Institute (EPRI), a French

participation including Framatome, Electricite de France (EdF) and

Commissariat a L'Energie Atomique (CEA), an Italian participation including

Energia Nucleare Ed Engeria Alternativa (ENEA), Ente Nazionaleper L'Energia

Elettrica (ENEL), and the Breda Division of Ansaldo Componenti S.p.A., and

participation of a Japanese consortium managed through the Nuclear Power

Engineering Test Center (NUPEC), including Mitsubishi Heavy Industries and

several Japanese utilities. The multi-sponsor nature of-the program contributed to a broad technical review of program activities, opened the program

to many technical, and technology inputs and provided a forum for the discussion

of steam generator issues on a world scale.

3.2 THE SURRY STEAM GENERATOR

3.2.1 Generator History

The Surry 2 steam generators had been in service only about 6 years at the

time of their removal. However, they exhibited, or had potential for

exhibiting, many of the defect types identified at the time the SGGP was being

formulated. The Surry nuclear station is sited near the brackish James River

near where it empties into Chesapeake Bay. The Surry Plant utilized 90:10

Cu:Ni alloy condensers that had a history of leakage during service. Water

chemistry records at the plant indicated periods of significant chloride

ingress into the secondary water. Oxygen ingress into the generator was also

not well controlled. The generator started operation in March of 1973 under

a phosphate secondary-side water chemistry. After .. about a year-and-a-half of,

operation, the secondary water chemistry was switched to an all volatile

treatment (AVT). The unit was subject to sludge lancing several times, includ3.2
ing an effort during the shutdown at the time of-water treatment-,switch-over.

However, eddy-current sludge height data indicate this mainly succeeded in

moving the sludge around as opposed to a significant. removal. During the

short period of phosphate water chemistry, wastage was fou'id and tubes plugged in the nozzle-side central tube region of the hot-leg (commonly referred to

as the 'banana zone').

Shortly after conversion to AVT water chemistry, the first indications of

denting were discovered. Initially, these were listed as unspecified

indications during the generator ISIs. The Surry generator has drilled hole

carbon steel support plates. The generator is tubed with mill annealed

Inconel 600. The acid conditions of the brackish water intrusions plus the

presence of chlorides and oxygen resulted in a rapid deterioration of the

Surry unit due to corrosion of the support plates. This corrosion led to

extensive tube denting, "hourglassing" of the support-plate flow slots and

fracturing of the support plates between tube ligaments. Additionally, the

hourglassing of the uppermost support-plate flow slots caused inward

deflection of the U-bend legs that resulted in high stress at the U-bend apex and initiated IGSCC which ultimately resulted in a large leak failure.

Following this event, the inner two rows of U-bends were preventively plugged.

Another feature that was of potential interest was the tube sheet crevice,

where the tubes have only been rolled into the bottom about 2-1/2 in. of the

21-in.-thick tube sheet. This extended crevice was the location of a "new"

degradation mechanism, intergranular attack of the tubing, at the time the

program was evaluating acquisition of a steam generator.

The Surry steam generator was judged suitable for the SGGP based on

several considerations. The unit was of a type found in a large number of

reactors; generators of this general design were licensed and produced in

Japan and France. The Surry unit was also believed to have defects

representative of those of concern at the time this program began. The Surry units were also the only generators of this type scheduled for removal in a

time frame compatible to the desired program start-up. A detailed report

describing the inservice history of the Surry steam generator was prepared by

Doctor et al. (1983).

3.2.2 Generator Acquisition and Transport

Prior to acquiring the Surry generator, an examination was conducted at

the Surry site to evaluate the generator from the secondary side. This

examination was to determine the condition of the generator after storage in

the vault at Surry. This examination, attended by representatives from NRC,

EPRI, Combustion Engineering, Westinghouse, Babcock and Wilcox, and Kraftwerk

Union, determined that the unit was in a condition typical of the service

condition at end of life. Thus, it was felt that no noticeable continued

deterioration had occurred during storage of the unit. A report by Wheeler

(1984) documented the results of this investigation. At The time of this

inspection, a small amount of water was visibly present in the generator

secondary side at the low point while the generator was on its side; however,

magnetite deposits had not changed to hematite. The examination at Surry

also ascertained that the service degradation did not preclude a successful

3.3
shipment to Hanford. Along with the visual inspection, and photographic

documentation of the generator, internal dimensional measurements were made

for post-shipment comparison. Repeat inspection of the unit after transport

to Hanford and installation in the Steam Generator Examination Facility. (SGEF)

indicated that no damage occurred during the transport (Wh;eler 1984).

The Surry unit was barged from the Virginia reactor site, through the

Panama Canal, to the Hanford Reservation in Washington State. Licensing for

the transport was pursued through the U.S. Department of Transportation. A

draft Environmental Impact Statement (EIS) was prepared and accepted for barge

transport of the generator as single item cargo. This report included data

from radiologic inventory measurements as well as risk analysis of various

accident scenarios. The generator shell'was accepted as a low specific

activity (LSA) package, based on the radioactivity being evenly distributed

throughout the tubing, and retained in place in the oxide coating on the inner

diameter of the tubes.

During shipment, precautions were taken to prevent ingress of oxygen and

water into the unit as well as to monitor and limit accelerations. Once at

Hanford, the unit was stored outdoors until the Steam Generator Examination

Facility was completed. While stored, the unit was kept under an inert

atmosphere by continuous feed and bleed of gas. The system was set up to

bleed in sufficient gas to prevent the entry of atmosphere into the unit during

normal daily thermal cycling. Shortly after storage, the unit was tapped

through the shell at its lowest point and several liters of water were drained

from the unit. This water had very low chloride content and apparently existed

in the unit prior to transport. It probably had drained into the unit from

the primary side of leaking tubes or was, in part, a result of condensation

during the barge trip. Figure 3.1 shows the steam generator as it arrived by

barge at Hanford.

3.2.3 Steam Generator Examination Facility

The Steam Generator Examination Facility, a cut-away drawing of which is

shown in Figure 3.2, was especially 'designed and constructed to meet the

anticipated programmatic needs of the SGGP. Initiation of SGEF construction

was delayed until after successful transport to Hanford of the Surry steam

generator. This sequential approach was necessary because of uncertainties

concerning the ability to license shipment of the generator. The SGEF was

more fully described by Clark and Lewis (1982). The main facility feature is

a 5-story high bay or tower with removable roof, which allowed positioning of

the generator in its normal vertical operating position. Vertical positioning

was judged crucial to the ability, to obtain representative nondestructive

examinations of the generator. It was also beneficial to activities such as

decontamination and some specimen removals. The tower portion of the SGEF

was environmentally [High Efficiency Particle Arrestor (HEPA)] filtered to

allow destructive sectioning operations on the generator. The remainder of

the SGEF contained a laboratory with filtered hoods, a truck lock/remote

operating area, change rooms and mechanical equipment to support the tower

operation. The SGEF met the needs of the SGGP; however, the tower portion was

somewhat cramped for optimum efficiency at positioning sectioning equipment.

3.4
IL

U

FIGURE 3.1. Surry Steam Generator Upon Arrival at Hanford

3.5
STEAM GENERATOR

I TON JIB CRANE

n. 4

Ill r-wN'iE

I ly

2nd LEVEL

(A

FIGURE 3.2. Cut-Away of SGEF Showing Position of Steam Generator
The facility could also have benefited from additional capacity to store

radioactive samples.

3.3 PREPARATION OF THE GENERATOR FOR RESEARCH

This section provides details on those tasks which were not central to

the primary objectives of the program, however were necessary precursors to

obtaining the data vital to accomplishing the primary objectives... These

activities included health physics, decontamination of the channel-head

region,land removal of plugs from tubes which were isolated during the

generator service life. While the tasks discussed in this-.section were

subsidiary to the major program goals, they were by no means trivial

accomplishments. Most of these activities often required first-of-a-kind

solutions to engineering and operational problems.

3.3.1 Health Physics, ALARA Radiation Control

One of the earliest and continui~ng concerns throughout the SGGP was the

potent-ial for large amounts of staff radiation exposure. While there was

continuing concern with minimizing staff exposure in general, there Was the

additional, potential in this program ofexcessiveiexposure burden on the

relatively few scientific staff pos'sessing the expertise to accomplish program

objectives. A separate Health Phys'ics task was established to accomplish

ALARA control of radiation'exposuresi' This task was responsible for reviewing

operating procedures from a radiati~on -control/exposure standpoint, overseeing

the accumulation of staff exposure information on'a time as well-as- job basis,

and assisting in experimental design for minimizing exposures, such as through

the use of selective shielding. Beyond the immediate needs of the SGGP, the

Health Physics staff sought information, in conjunction with another NRC

program (Reece and Harty,'1983), to aid in accuratelymonitoring and reducing

worker exposures in the nuclear industry.

Initial Health Physics activities included a scan at the Surry site to

determine the radionuclide inventory. The inventory was estimated to be

between'120 Ci and 180 Ci, composed almost exclusively of Co-60. These

radiologic data, were used during the shipment licensing activities, and as adesign basis for transport-shielding. A health physicist accompanied the

barge shipment on the tow boat, in case there was a necessity to provide

radiation monitoring upon putting into a port. This was indeed the case when

entering the Panama Canal. '

Once the generator was lowered into the SGEF, the first subsequent effort

was establishing and mapping the radiation fields inside and around the SGEF.

The inside'maps, an example is provided in Figure 3.31 allowed the design of

shielding,.the training of personnel,, and experimental activities/setups, to

be conducted in a manner which minimized staff radiation exposure. The

outside mapping was the beginning of an environmental monitoring program which

lasted throughout the SGGP. This effort included location of dosimeters at

10 points surrounding the facility perimeter. Periodic monitoring of these

environmental dosimeters exhibited a close correlation to activities inside

the SGEF, such as decontamination, and removal and storage of artifacts in

different portions of the facility. Based on environmental monitoring,

3.7
LEVEL - 1st FLOOR LEVEL - 2nd FLOOR

-*DOSE RATES ARE IN mR/hr *DOSE RATES ARE IN mR/hr

LEVEL - 3rd FLOOR LEVEL - 4th FLOOR

*DOSE RATES ARE IN mR/hr

FIGURE 3.3. Radiological Maps - Floors 1-4 SGEF
additional concrete shielding was placed outside the SGEF when needed to

maintain perimeter exposure levels well below legal limits.

The Health Physics task was also charged with developing an on-line

dosimetry system to keep close track of personnel exposures. Each entry into

the radiation zone was monitored by individual pencil dosimeter, and logged

into a computer data base by a radiation protection technician assigned to the

facility. All personnel using the SGEF were subject to radiation training,

building-specific training, and task-specific training. This resulted in an

extremely low accident rate and personnel contamination incidence during the

program.

In addition to mapping around the generator, the radiation fields throughout the inside of the generator were also mapped, both using thermoluminescent

detectors (TLDs) through shell penetrations and utilizing a special probe

through the inside of the generator tubes. Maximum dose rates in excess of

11 R/h were recorded near the generator center. The through-tube measurements

were conducted using a specially constructed single crystal 0.5-in. diameter

CdTe probe. The detector was pushed through 48 selected tubes using a NDE

probe pusher-puller. Figures 3.4 and 3.5 show typical hot-leg and cold-leg

detector scan results. Note that the support-plate locations are discernible.

The relative radiation level increases from the inlet of the tube to the

U-bend, and remains at the maximum value throughout the cold-leg side. This

is typical of all of the tubes that were scanned. It is probably a result

of increasing deposition of Co-60 along a tube as primary water temperature

decreases (decreasing solubility of iron and cobalt). The radiation levels

were consistently higher in the manway quadrant than the nozzle (inlet)

quadrant of the hot-leg. Again, this is interpreted as resulting from

decreasing solubility, owing to lower temperatures and, perhaps, less

turbulence. On the cold-leg side, the levels were equal on the manway and

nozzle (outlet) quadrants.

The internal radiation scan was an unsuccessful attempt at seeing if such

measurements could be used to detect defects in the generator tubes or support

structure. Eight tubes with eddy-current indications of 50%-100% throughwall defects were scanned. The negative finding was due to the fact that the

contribution to the radiation field of the tube being inspected is small

compared to the surrounding tubes. The radiological scans do appear to allow

a good thermal map (i.e., relative operating temperature) of the generator to

be developed (Clark and Lewis 1985).

Another activity under the health physics task, and in conjunction with

another NRC program (Reece and Harty 1983), assessed the multiple dosimetry

of channel-head workers (how many dosimeters are needed and where they should

be located on an individual in order to ensure adequate dosimetry). The

results of this study (Wheeler et al. 1984; Clark and Lewis 1985) was that

exposure to the eye was the limiting exposure. The recommendation was that

channel-head workers were adequately monitored with two dosimeters, one

located on the forehead and the other on the chest.

3.9
S/'G TUBE R22 C47 HOT LEG

0 SEC. LIVE TIME 101 SEC. TRUE TIME

3200

2400

+ +

.4

+

CD)

71

0

U

160.0

800

0

F

I I I I , I 

, 

I 

I

I I 

I I i 

I I,

I I I I

0 256 512

CHANNEL

768 1024

FIGURE 3.4. Relative Radiation Level Along a Typical Hot Leg Tube - R22C47
S/G TUBE R22 C47 COLD LEG

0 SEC. LIVE TIME

3400

2550

i~i SEC. TRUE TIME

s 4 4

*

_J

z

C-)

.,)

F-.

z

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0

U._

4

4

1700 FI

4

4

850 -

i

*

4

I I I I I i I I I I' I f I f I

0

0 256 512

CHANNEL

768 1024

FIGURE 3.5. Relative Radiation Level Along a Typical Cold Leg Tube - R22C47
3.3.2 Channel-Head Decontamination

This task was included primarily to maintain staff radiation exposure on

an ALARA basis. Access through the channel-head was needed for tube

unplugging, primary-side nondestructive examinations, removal of specimens,

and demonstration of generator repairs. The programmatic goal of the

channel-head decontamination was to reduce the general area radiation level

in the channel-head to below 1 R/h, without decontamination of the tubing.

The field practice technology for channel-head decontamination at the

time of this activity was abrasive blasting using a robotic or semi-robotic

manipulator system. The technique had several drawbacks such as: generation

of substantial secondary waste, incomplete coverage with most manipulators,

and staff exposure during installation of necessary equipment. The SGGP

solicited for subcontractor bids to perform the channel-head decontamination.

During the technical evaluation of the proposals received, it was decided to

allow demonstration of two techniques, one on either side of the channel-head

divider plate. The two techniques both utilized dilute reagent chemical

decontamination; a CAN-DECON and a LOMI method were chosen. Both of these

techniques had prior extensive laboratory data on decontamination effectiveness

and potential corrosivity. Each technique had prior demonstrated success in

boiling water reactor application. The contractors offering the techniques

had previous radioactive field experience. Interest in the dilute chemical

reagent decontamination techniques stemmed from their ability to cover all

channel-head surfaces, to potentially avoid personnel exposure during setup

and operation, and to provide reduced generation of secondary waste through

the use of ion exchange water cleanup. Due to concerns regarding possible

corrosivity, it was also felt these techniques required a full-scale

demonstration before they could be used for the first time in the field.

A report by Allen et al. (1984), provides details on the execution and

results of the decontamination efforts. This report contains information on

the equipment and operations, corrosion tests conducted and radiologic

measurements. In summary, both techniques achieved average surface

decontamination factors (DF) up to a factor of 10, with higher decontamination

on stainless steel surfaces and DFs of 4-6 for Inconel 600 surfaces. Both

techniques were limited to decontaminating the channel-head bowl and bottom

surface of the tube sheet. Decontamination into the tubes was prevented, to

avoid any changes to the service-induced tube degradation conditions in the

tube sheet region, and to avoid possible spillage of chemical decontamination

solutions into the secondary side. Thus, after decontamination, considerable

radiation field shine down the tubes remained. Fields in the channel-head,

after the decontamination, averaged less than the program goal of 1 R/h.

Both decontaminations were continuously monitored with resistance type

corrosion probes. In addition, numerous corrosion coupons- U-bend strain

specimens, and metal couples were exposed to the solutions for post-process

evaluation. The corrosion test results indicated essentially no detrimental

effects for the period of exposure. It was also interesting to note that

on-line process monitoring of radionuclide concentrations.(see Figures 3.6

and 3.7) indicated that process times could probably be reduced, and provided

detail on how the decontamination process proceeded.

3.12
40

30

z

0

4

z 20

0

C.)

1

o 1

U,

U

M

z

0

z

0

U4

2

2

0

30

z

2

U

z

0

U

0

0

U)

U)

>,

20

- Fe

- - Mn

- - - - Cr

- -r

I I C 1....

10

0

0 40 80 120

TOTAL ELAPSED TIME (hr)

160 200

FIGURE 3.6. Fe, Cr, Ni,

the Reagent

Mn, and Co-60 Concentrations in

During the Hot Leg Decontamination

3.13
30

z

20

U

z

0

U

40

o i

03

z

U,

z

0

20

x 0

SO

40

FIUR 307

z

0

0

,-m

X

ca

0 24 48 72 96 120

TOTAL ELAPSED TIME (hr)

Fe, Cr, Ni, Mn, and Co-60 Concentrations in the Reagent

Before Ion Exchange During the Cold Leg Decontamination

3.14
One of the processes demonstrated the ability to clean waste streams

with ion exchange beds, providing for reduced volumes of secondary waste

generation. One problem caused by the ion exchange was very high fields around

the exchange columns, which necessitated additional personnel shielding and,

in the case of the SGEF, additional facility shielding to maintain low

radiation field values at the facility perimeter. The high concentration of

radioactivity in a small portion of the ion exchange media, required some

dilution to achieve fields of 3-5 R/h at surface contact on the radioactive

waste disposal drums. The ion exchange media was buried at the disposal site

at Hanford.

The main lessons demonstrated with this task were 1) dilute chemical

decontamination as conducted in the Surrv unit provided average field

reductions approaching a factor of 10; 2) the processes used did not produce

appreciable corrosion damage during implementation; 3) secondary radioactive

waste generation could be minimized with use of ion exchange solution cleanup;

and 4) operating personnel exposures were low, 1.6 R for the initial

decontamination side, and 0.7 R for the second side of the channel head.

Other observations included: 1) Decontamination effectiveness was

enhanced by proper design of the system to provide reasonable flowing

circulation over all the surfaces to be decontaminated (this is a function of

flow recirculation rate and in-channel-head solution distributor or nozzle

design); 2) Post-decontamination high-pressure water spray removed residual

loose deposits, though did not account for significant radioactivity

reduction; and 3) A semi-remote electropolishing demonstration removed

remaining corrosion films down to bare metal.

3.3.3 Tube Unplugging

During the approximately six years the Surry 2A generator was in service,

748 of the generator's 3388 tubes, just over 22%, were plugged. Plugging

removes a generator tube from the pressurized primary water circuit. In a

U-tube type steam generator, such as Surry, this involves closing the tube at

the water entry and exit points, by obstructing it within the tube sheet

region. The Surry tubes were plugged using an explosive plug, a cutaway

section of which is shown in Figure 3.8. Tubes were plugged for a variety of

reasons during the lifetime of the Surry unit. Reasons included: identified

and unidentified types of tube wall degradation, preventive plugging based on

engineering judgment to preclude tube failures from support-plate degradation,

and preventive plugging of the inner two rows of U-bends after a U-bend burst

failure. Figure 3.9 shows the end-of-service Surry 2A generator tube plugging

map, provided by the Virginia Electric and Power Company.

The SGGP required access to a large number of the plugged tubes for

nondestructive inspection. Tubes plugged due to wall-loss/degradation were

of primary interest in the program efforts to validate NDE accuracy and

reliability. Figure 3.9 indicates tube plugs to be removed, on both hot-leg

and cold-leg sides. All tubes plugged for a wall degradation, a defect or

leaker, were unplugged. In addition, half the preventively placed plugs, on

one side of the tube sheet, were removed along with a selection from the

opposing side to confirm symmetry of generator degradation.

3.15
*1

932 £v;Ž~w <-K'

~t%' <~xhu

V

FIGURE 3.8 Sectioned Explosive Plug in Simulated Tube Sheet

3.16
Color-Plugging Reasons 7".(!0066666&TW7 45 7778CM 866666 oG ??? 4777

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-NOZZLE MANWAYVEPCo Tube Plugging Reason Map

FIGURE 3.9. Surry 2A Steam Generator Tube-Plugging Map as of 8/8/78
The removal effort itself was a complex organizational activity. At the

time, this was the largest plug removal operation undertaken in the U.S.

Previous experience was limited to removals of one to a few plugs at a time.

This project required the removal of close to a thousand plugs. Plug removal

activities were documented in a report by Wheeler et al. (1984).

Removal of the plugs shown in Figure 3.8 involved a three step process.

First, a small hole was drilled through the solid end of the plug. This hole

released any trapped fluid above the plug and provided a gripping orifice for

later stub removal. Next, the plug body was drilled out using a drill bit

that matched the tubing inner diameter. Drilling proceeded until the solid

top became free-moving. Finally, a special tool was inserted through the now

loose plug end, and the end removed. The majority of removals utilized

equipment diagrammed in Figure 3.10. The plug to be removed was indexed by a

fixture attached to the tube sheet. To this fixture was attached a reversible

air-powered drill motor with an electric feed. Channel-head workers,

"jumpers," would attach the drilling equipment and change bits as needed.

The trained drill operators used the equipment from a remote location with

visual and audio inputs from the drilling operation. Using this technique

along with two simultaneously operating drills, one on either side of the

channel-head, plug removals in excess of 50 per day were routinely

accomplished.

Containment of contamination and personnel exposure control were concerns

during the unplugging operation. In addition to metal turnings from the

drilling operation, a number of tubes contained water behind the plugs.

Channel-head workers were required to wear two pairs of protective clothing,

plus a water repellent suit, and leather gloves to prevent cuts from the metal

turnings. Workers operated for 2-hour periods with a 2-hour break between

periods. Pressurized breathing air was utilized, which also provided some

cooling. In addition, the SGEF was air conditioned. The limiting

channel-head worker radiation exposure was to the head/eyes. Exposure rates

typically varied between approximately 25 mR/h and 115 mR/h. The exposure

variation was a function of the number of channel-head entries made per work

period and their duration. Duration was affected by any mechanical problems

with the equipment, worker experience, and the difficulty of positioning the

indexing device and tooling, which was a function of the plug location,

especially for peripheral plugs. Total personnel exposure for the unplugging

operation was 60.6 man-rem with channel-head workers accounting for

40.3 man-rem.

Several tubes which had been preventively plugged, and other tubes which

were not leakers at the time of plugging, i.e., had no through-wall defect,

were found to contain water and/or sludge deposits. Figure 3.11 shows the

distribution of the water and sludge found while unplugging tubes. Water

samples were obtained from 71 of the 102 plugs visually observed to leak

during unplugging. Analysis of these samples indicated that most of them

contained water typical of secondary-side chemistry, while a few indicated

primary water. Water origin was identified mainly through high lithium and

boron levels for primary water, and high copper, sodium and phosphorus for

secondary water.

3.18
MOTOR & LEAD SCREW

LOCKING DEVICE TUBE SHEET WITH

BASE PLATE

CAMLOCK \

CAMLOCK PLATE

PLUGGED TUBE

'0000000

0000000

0o000009

I-.

i.0

DRILL

AIR DRIVEN

DRILL MOTOR

LEAD SCREW

ASSEMBLY

FIGURE 3.10. Schematic of Model D Plug Drilling Fixture
46

452

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42 41

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COLUMN NO

eO1--NOZZLE MANW -- wMW D

2A STEAM GENERATOR NOTES

ITUBE SHEET VIEWED FROM INSIDE PRIMARY PI (HUMl WITH OBSERFVERFS BACK TO DIVIDER PLATE) 1 TOTAL NUMBER OF TUBES 3.383

2 aCOORDINATE 1.281 3 yCOORDINATE 1,231

4 TUBE 0 0 73a

Red = I-ater 5 TUBE WALL . 060"

Yellow = Sludge

FIGURE 3.11. Location of Water and Sludge Samples
The drilling operation overall was very successful. A total of 969 plugs

was removed duringa 3-week period. Toward the end of the operation,

productivity was affected due to equipment wear requiring increased

maintenance. There was also a problem due to the drilling. After drilling,

the tubes were mechanically gaged to determine that the drill hole was

correctly sized. This gaging was mainly conducted so that "finger-walker"

type eddy-current probe positioning equipment could later be used. The tube

sheet holes had to be within a certain size tolerance for the finger-walker

to maintain an adequate grip to prevent falling. However, the gaging operation

was only to a depth affecting the finger-walker. What was missed were

occasional ridges in the steam generator tubing left at the end of drill

penetration, due to drilling at a slight angle (caused by misalignment of the

indexing device) or when the drill was off center (caused by variation in the

tube sheet hole spacing). These ridges later caused probe hang-up and damage

during eddy-current inspection. Fortunately, the problem was limited to a

few tubes.

3.4 NONDESTRUCTIVE INSPECTIONS

Following the preparatory tasks, the steam generator was examined visually

and nondestructively to characterize the number, type and distribution of

tubing defects. This section describes the'results of visual inspection of

the secondary side alongwith the two base-line EC inspections. Data

developed from these inspections were used to select a subset of 320 tubes

for the round robin activities which are described and summarized below.

3.4.1 Secondary-Side Examination

Examination of the Surry generator from the secondary side was conducted

with several objectives in mind. These inspections provided information about

the tube bundle support structure. They served as an alternate inspection of

the generator tubing, providing comparison to the data derived through primaryside inspections. The secondary-side effort also provided sampling and

assaying of materials and corrosion products and deposits. Another purpose

was to see if anything could be determined concerning generator flow patterns,

based on observing tubing films, coloration and sludge accumulation.

Secondary-side inspections were also being discussed and under consideration

by NRC for field application. The SGGP effort was partly aimed at determining

if the inspections could be conducted, and what technology was available or

needed to achieve useful data.

3.4.1.1 Inspection Techniques

Initial secondary-side characterization was conducted through the three

relatively large 12-in. square penetrations torch-cut into the generator for

the preshipment examination, along with the two 5-in. diameter handholes

located between the top of tube sheet and the first support plate. Two of

the preshipment penetrations, at the first and seventh (uppermost) support

plates at the tube lane, provided reasonably spacious access for examination

of the top of tube sheet and inner row U-bends, respectively. Characterization

in these areas was conducted with a miniaturized 35-mm camera, modified for

remote triggering, and with rigid boroscopes. Later, the outer row U-bends

3.21
were accessed through removal of the cover plate that closed the tube bundle

portion of the steam generator. This region was also examined with the use

of a periscope (Schwenk and Wheeler 1984).

None of the above techniques provided for viewing a significant distance

into the tube bundle. The bundle consists of 3388, 0.875-in. diameter tubes

in square array located on 1.28-in. centers, leaving a nominal 0.405-in. space

between tubes. This space was usually further reduced due to corrosioninduced damage. Fiberscopes were employed to examine the inside of the tube

bundle. While the fiberscopes did provide some useful secondary-side

characterization information, they suffered from a number of shortcomings.

Major problems resulted from the limited viewing area and great difficulty

with remotely positioning the fiberscope. Knowing where an observation was

being made (i.e., tube row and column) was quite difficult, as well as

relocating a previous position.

To overcome the difficulties in performing these inspections within the

often deformed tube bundle, the SGGP developed and used a series of miniature

pinhole cameras. This technique was enormously successful in allowing

photographic documentation virtually within any region of the tube bundle. A

relatively large region could be photographed in a limited time, important

for minimizing personnel radiation exposure. The minicameras also facilitated

documenting the exact location of examination. The photographic techniques

for secondary-side characterization are discussed in detail by Sinclair (1984).

Figure 3.12 shows typical pinhole cameras. Figure 3.13 provides an example

of the photographs obtained.

Access to the secondary side of the tube bundle was accomplished through

multiple shell penetrations made using a 2-1/2-in. diameter rotabroach.

Typically, penetration of the 2- to 3-in. thick generator shell could be made

in approximately one hour, including tooling setup and drilling. Where

necessary, overlapping penetrations were made, to provide larger access areas.

Shell and bundle shroud penetration techniques had to be nondamaging to the

tube bundle; this eliminated such thermal techniques as plasma arc or

oxyacetylene torch. Electrodischarge machining was eliminated from

consideration because of the need for large amounts of liquids that, during

breakthrough, could enter the generator and possibly alter the character and

disposition of sludge and other corrosion products.

3.4.1.2 Inspection Results

The majority of the secondary-side inspection work was conducted prior to

completion of the primary-side nondestructive inspections. As such, the

secondary-side inspections, and acquiring access for the inspections, had to

be nondamaging to the tubing.

3.4.1.2.1 U-Bend Area

The inner two rows of U-bends were preventively plugged, following a

sudden U-bend failure during service. Stress corrosion cracking of U-bends was

affected by tube support-plate degradation and by fabrication-induced

stresses. The secondary-side inspection revealed that as many as 20 Row 1

3.22
FIGURE 3.12. Cameras Fabricated for Use in Surry Generator
FIGURE 3.13. Side Movement (Parallel to Lane) of Tubes Relative

to the First Support Plate, Column 39-40

3.24
U-bends showed burst type cracks. In addition, 50 to 66 other Row 1 U-bends

were probably cracked either on the extrados side of the U-bend or showed

axial creases on the intrados side, probably indicative of primary-side

initiated cracks that had penetrated part way through the wall. Figure 3.14

shows a creased tube.

FIGURE 3.14. Probable ID (Primary) Side Initiated Crack Which

Appears as an Axial Crease in the Apex of Column 87

All inner row U-bend failures were located in tube columns which

corresponded to support-plate flow slot "soft spots." Figure 3.15 reveals

support-plate flow-slot hourglassing in the seventh support plate.

In general, visual and photographic inspection did not reveal cracks or

shallow defects beneath the thin, brown-black outer scale found on most tubes.

Instances of missing U-bend sections, such as shown in Figure 3.16c, were

revealed. Since the inner two rows of U-bends were plugged after an initial

U-bend failure, the numerous through-wall defects, and the missing tube

sections, occurred after plugging. The cracking could, in part, be

attributable to the increasing applied stresses as the support plates continued

to corrode. No other U-bend damage was observed via the in situ secondary-side

inspections.

3.25
-Jkt-~1 Ah('% Column 94 t

Column 46

wA

Column 48 Column 1

8605834-4cn

FIGURE 3.15. Montage Photo of the Top of the Seventh Support

Plate at the Tube Lane, Column 94 to Column 1
a) b)

c)

FIGURE 3.16. Probable Burst-Type Fracture, U-Bend R1C85;

a) Penetration Below Seventh TSP, b) Looking Up

(Intrados Region), c) Looking Down (Extrados Region)

3.27
3.4.1.2.2 Tube Support Plates

Corrosion of the tube support plates leading to tube denting and

support-plate damage (cracking and deformation) was the primary cause of tube

plugging in the Surry generator, eventually leading to its early retirement

from service. Secondary-side inspections documented support plate conditions

throughout the generator and concentrated on an examination of the second

support plate (Schwenk 1987). The second support plate was considered the

most severely degraded based on eddy-current and profilometry testing.

Tube lane flow slots at each of the seven support plates were severely

deformed. The traditional hourglassing of the flow slots was only seen on

the uppermost support plate. At all other support-plate levels the

hourglassing had progressed such that the flow slots were essentially closed

due to deformation and collapse from the hot-leg side, such as shown in

Figure 3.17. The detailed secondary-side examination of the second support

plate was to determine how extensive damage was across the support plate and

the potential effect on remaining structural integrity. A particular concern

was how far back from the collapsed flow slots did significant tube damage

occur. Figure 3.18 documents the support plate condition completely across

the hot-leg side at one column. Several regions of the second tube support

plate were documented in this manner (Schwenk 1987).

Flow hole compression/ovalization was documented near the expected support

plate hard spots between flow slots and at block/wedge locations anchoring

the support plates to the generator shell. Extensive flow-hole-to-tube-hole

ligament cracking was noted, though this condition could not be fully

photographically documented due to sludge/corrosion product accumulation on

the support plates. Sections of support plate were found as loose parts

within the generator.

After removal of the U-bends, full-length tube pulls from the top of the

generator experienced little or no resistance on the hot-leg side, despite

the expected compressive loadings anticipated from extensive corrosion and

denting of the support plate structures into the tubing. The support plates

on the hot-leg side readily broke up, indicating ligament failure across the

entire hot-leg side. The cold-leg support-plates, especially those below the

fourth support plate, did not readily break up during specimen removals

(although they exhibited significant corrosion).

3.4.1.2.3 Top of Tube Sheet/Sludge Pile

The Surry generator experienced wastage damage during initial service, using

a coordinated pH phosphate water chemistry control. Sludge lancing was

conducted; however, eddy-current sludge height measurements indicated sludge

was moved around on the top of tube sheet, but not effectively removed. The

sludge pile was found to consist of two distinct regions, a granular black

sludge (AVT origin) on top of a layer of phosphate sludge. The pile was

located mainly in the center portion of the tube bundle; however, there was

also sludge in the peripheral regions. The sludge in the peripheral regions

was on thp tubesheet and attached to the tubes. The phosphate sludge was

hard and somewhat brittle. Sludge samples were acquired from tube sheet

3.28
A ý6L 44

I I ~i)

V

FIGURE 3.17. Bottom Side of First Support Plate, from 00 Handhole
Lj

0

FIGURE 3.18. Montaqe Photo Lookinc Up at the Second Support Plate from

Shell Wall (Hot Leq Side) to the Cold Leq Side at Column 62-63
section removals, as shown in Figures 3.19 and 3.20. Sludge composition varied

considerably with height above the top of tube sheet (TTS) and from the hot-leg

to the cold-leg side of the generator. There were also compositional

variations from point to point in a local region. Tables 3.1 and 3.2 provide

a summary of sludge analysis as a function of location above the TTS for

hot-leg and cold-leg regions. Results from inductively-coupled plasma

spectroscopy (ICP) are presented in Table 3.1. Iron, copper, and zinc were

the primary elements found in the hot-leg sludge samples. The Cu and Zn

concentrations increased with distance above the TTS with a sharp increase in

copper observed in the AVT sludge. In addition to Fe, Cu, and Zn, high concentrations of Na and P were found in the cold-leg sludge samples from locations up to 1.5 in. above the TTS. A sharp increase in Cu and Zn and corresponding decrease in Na and P occurred when the sludge changed from phosphate

to AVT origin.

X-ray diffraction (see Table 3.2) analysis indicated that AVT sludge was

primarily composed of Fe30 4 and metallic Cu with a small amount of Fe 203 in

the cold-leg sample. The reddish phosphate sludge from the cold-leg sample

was composed of Fe20 3 , Fe3 0 4 , and NaFeP0 4 in decreasing order of relative

concentration. Only a trace of Fe 30 4 was detected in the X-ray diffraction

patterns of the phosphate sludge samples taken from near the TTS of both the

hot- and cold-leg sludge piles. A lack of diffraction peaks in these samples

suggests either an amorphous or highly defected crystal structure with poor

diffraction characteristics. Further information on sludge characterization

may be found in reports by Kurtz et al. (1987) and Bradley et al. (1988).

3.4.1.2.4 General Secondary-Side Observations

The secondary side was characterized throughout the generator. In

general, the tube-support structure appeared to be heavily corroded, cracked

and deformed. This deformation often caused tube damage a considerable

distance (-6-8 rows) back from boundaries between hard- and soft-spot tube

support-plate regions. Several examples were documented of loose parts in

the generator, consisting mainly of small sections of support plate. There

was no positive association of tube damage with the loose parts. The loose

part locations were used as inputs to the test matrix, to ascertain if the

primary-side NDE would detect and define the resulting signal. Tube

support-plate sections removed from the hot-leg invariably fell apart,

indicating complete loss of structural integrity in the support plates.

Tube bursts in the inner row U-bends, as well as bulged tubes, were found.

It did not prove feasible to attempt definition of fluid flow within the

generator through assessing color changes on the outer film of the generator

tubes. The main color differences noted were between plugged and nonplugged

tubes, and these differences were mainly apparent within or near the sludge

pile.

The most comprehensive summary of the secondary-side inspections is found

in a report by Schwenk (1987). This effort did prove that the secondary side

of a generator can be inspected. The resulting documentation is the most

comprehensive ever published on the conditions which can derive from corrosion

damage wit'in a steam generator. The secondary-side characterization also

3.31
A>A

FIGURE 3.19. Overview and Sludge Pile Closeup of Hot Leg Tube

Sheet Section #355

3.32
FIGURE 3.20. Overview and Sludge Pile Closeup of Cold Leg Tube

Sheet Section #354

3.33
TABLE 3.1. Results From ICP Analysis of Sludge Samples From

Tube Sheet Specimen #355 (HL) and #354 (CL)

Location Above Tube Sheet, mm

0-10

Element HL CL

10-20

HL CL

25-35

HL CL

Fe

Cu

Zn

Cr

Ni

Si

P

Na

67.00

0.81

0.28

0.26

0.62

0.26

0.29

0.15

35.05

1.44

0.27

0.29

0.49

0.73

9.77

12.90

55.41

3.76

1.05

0.25

0.89

1.06

1.35

0.13

34.69

3.20

0.39

0.13

0.33

0.24

10.57

13.35

42.53

18.07

2.24

0.27

1.19

0.76

1.25

0.17

36.28

4.57

0.57

0.26

1.07

0.46

10.08

8.70

50-70

HL CL

40.17 44.51

21.90 14.46

6.94 11.27

0.14 0.20

0.12 0.81

0.34 0.49

0.31 0.24

0.80 0.96

Values reported as weight percent.

TABLE 3.2. Results From X-Ray Diffraction of Sludge Samples

From Tube Sheet Specimen #354 (HL) and #355 (CL)

Location Above Tube Sheet, mm.

Phase

Fe30 4

Fe2 0O

Cu

NaFeP04

0-10 10-20 25-35

HL CL HL CL HL CL

(a) (a) (a) (a) 60%(b) 30%

.. .. .. . -- 50%

........ 40% -0-

.......-- 20%

50-70

HL CL

60% 25%

-- 15%

40% 60%

(a) Mostly amorphous with trace of Fe30 4 .

(b) Relative phase concentrations were estimated from peak intensities.

3.34
provided valuable inputs into the selection of specimens for primary-side

nondestructive inspection and for subsequent specimen removal to va~lidate NDE

and assess generator damage.

3.4.2 Base-Line Eddy-Current Inspection

The goal of this task was to characterize the generator throughout to

establish the total defect conditions observable with eddy-current examination.

The results ,of this characterization wouldthen be amenable to statistical

interpretation. This allowed a representative subset of specimens which were

characteristic of conditions found in the steam generator to be chosen for

the round robin examinations.

Prior to the post-service base-line examinations, an eddy-current

profilometry examination (using a*Zetec MIZ 15) had been conducted of 96 tubes

selected to represent all regions of-the.generator. 'This examination was

used to specify the probe sizes required for the base-line study, and hadshown that the generator could be reasonably inspected with large-diameter

(0.700-in.) probes over most regions. It also indicated that the location of

maximum denting was at the hot-leg second support plate (Clark and Lewis 1984).

The extent of denting and the effect on inspectability was a major preliminary

concern since the Surry 2A tube plugging was largely denting related.

Two post-service base-line eddy-current examinations were conducted. One

examination used the Zetec MIZ-12 multifrequency-eddy-current data acquisition

system with a Zetec DDA-4 data analyzer, which was the current field practice

standard in the U.S. The other examination utilized an Intercontrole IC3FA

multifrequency data acquisition system and an Intercontrole IC4AN data

analyzer. It was felt that these two systems and their close derivatives

account for the vast majority of current field inspections of nuclear steam

generators. Two examination teams were chosen to conduct the post-service

base-line examinations. Each team consisted of experienced ISI practitioners

and were expert with their respective system. Both examination teams were

requested to conduct the best possible examination, meeting as a minimum the

requirements set forth in the ASME Boiler and Pressure Vessel Code, Sections

V and XI. In addition, all wall-loss indications, including those less than

20% through wall, were requested to be reported. Both teams were aware of

the research nature of the project, and each team included senior analysts

from the respective vendors.

The base-line eddy-current examination of the steam generator often

involved repeated scans with progressively smaller probes, to obtain access

through dented support-plate regions. Probe diameters down to 0:610 in. were

used by Zetec and down to 0.669 in. by Intercontrole.

Comparing the results of the two post-service base-line eddy-current

examinations produced a spread in results greater than anticipated by the

project staff. There was disagreement on defect sizing, on classification of

eddy-current indications as a defect or as an artifact and, in a few cases,

on the existence of an eddy-current signal. Many of these disagreements

remained even after both teams were informed of specific differences and

allowed aa, opportunity to modify their respective results. In a few cases,

3.35
the teams admitted an omission or reinterpreted their results. Both teams

reported that the effects of denting and conductive deposits on eddy-current

signals could not be eliminated by mixing the responses of various test

frequencies. The differences in inspection results appear to be due to

differences in the analysts' interpretations of complex eddy-current signals.

As a. result of the data review, the teams' analysts agreed to disagree in

their interpretation of the data. Since the project could not determine a

priori if one or the other base-line inspection team was correct, the selection

of a subset of tubing specimens representative of the generator condition

became more complex.

The results of the base-line inspections, summarized in Table 3.3, show

that the majority of reported defects (wall-loss indications) were at the top

of the hot-leg tube sheet in the sludge pile. There was a-relative paucity of

defects in other regions of the generator, except for some indications in the

U-bend area not generally associated with AVB contacts. The base-line

inspections did not indicate degradation in the tube sheet crevice, though

some denting was present at the TTS. The base-line inspections also indicated

the presence of tube bulging in previously plugged tubes. The only common

detections between the two base-line inspection teams were at the top of tube

sheet on hot-and cold-legs, the first support plate on the cold-leg and the

seventh support plate on the hot-leg.

TABLE 3.3. Locations of Wall-Loss Indications

From the Base-Line Inspections

Number of Indications

Hot-Leg Cold-Leg

Reference Location Team X Team Y Team X Team Y

Tube End 1

Tube Sheet Gap 1

Unspecified Tube Sheet 2 4

Top of Tube Sheet 713 731 22 113

Support Plate 1 3 5 1 6

Support Plate 2 3 2

.Support Plate 3 2 3 2 1

Support Plate 4 1 .2 3

Support Plate 5 1 9 1

Support Plate 6 20 1 9

Support Plate 7 11 115 20 10

U-bend 22

Further details on the base-line eddy-current inspections and analysis

of results can be found in a report by Doctor et al. (1988a).

3.36
3.4.3 Round Robin Tube Selection

A major goal of the SGGP was to establish the reliability of eddy-current

inspection of PWR steam generator tubing. The two base-line inspections of

the Surry 2A generator were planned as a means to locate indications for

selecting a subset of tubes for NDE round robin testing and to supplement the

ISI reliability data base. The criteria for selecting an appropriate subset

of tubes from the Surry generator were based on all available information on

the generator, historical and derived by the SGGP. The first consideration

was to ensure that sufficient numbers of defects were included in the sample

to allow a statistically valid estimate of the ability of eddy-current

procedures to detect and size defects.

There was information on the presence and locations of possible defects

from two basic sources: 1) the inservice history of the generator (leakers,

plugging reasons) and secondary-side and other observations made during the

SGGP (defects, sludge-filled tubes at unplugging), and 2) the SGGP base-line

eddy-current inspections. The historical data did not provide much information

on possible wall-loss defects except in the crescent zone region near the

center of the hot-leg top of tube sheet. The eddy-current inspections

conducted in the two years prior to decommissioning were used only to determine

the extent of denting. Some through-wall defects were reported because of

observed leakage. Therefore, the majority of information on tubes containing

possible defects was taken from the post-service base-line inspections.

A statistical sampling rationale was derived as explained by Doctor et al.

(1988b). The sampling rationale involved a stratification of the base-line

inspection results into wall-loss indication size catagories, <20%, 20%-40%,

>40% of wall, and into three agreement catagories:

° Both teams reported an indication at the same location and agreed on the

depth within 10%.

* Both teams reported an indication at the same location and differed in

reported depth by >10%.

* Mismatch on detection - either the teams differed on the source

(reportable indication versus artifact) of the signal they both detected,

or one team did not detect a signal found by the other.

Each agreement category (strata) was sampled equally and the size strata

were sampled such that the final defect set contained 45%, 35% and 20%,

respectively, for wall-loss indications of >40%, 20%-40%, and <20%. Sampling

within the strata was done on a random basis. Eighty percent of the tubes,

-240, were chosen because defects were reported by the base-line inspection

teams. The remaining 20% of tubes selected for the round robin sample were

chosen for several reasons. They included tubes in which 1) a defect was

expected, but none was reported by the base-line inspections (i.e., surrounding

tubes all showed indications), 2) tubes in which no defects were expected

and none were reported, and 3) tubes in which non-wall-loss conditions were

reported (bulging, permeability variations, conductive deposits and loose

parts).

3.37
Table 3.4 shows the number of tubes selected for the round robin subset

in each category. Figure 3.21 illustrates the location of the round robin

tubes across the generator bundle.

TABLE 3.4. Numbers of Tubes in the Round Robin Sample

Listed by Size and Agreement Substrata

Indication Matched Detection Matched Detection Unmatched

Depth Size Diff <10% Size Diff >10% Detection

>40% 35 36 32

20%-40% 30 30 29

<20% 16 16 16

Miscellaneous Tubes Added to Matrix - 80

Total Round Robin Tubes - 320

3.4.4 Round Robin Inspections

A series of round robin nondestructive examinations were conducted on the

generator. These examinations were divided into:

* Data Acquisition and Analysis Round Robin (DAARR) using Zetec multiýfrequency EC inspection equipment

* Advanced/Alternate Techniques Round Robin (AATRR) consisting of NDE

inspections with equipment or analysis procedures significantly different

from those used in the base-line or other round robin inspections

" Analysis Round Robin I using EC data tapes produced by the base-line

inspection conducted with Zetec multifrequency equipment

* Analysis Round Robin II using EC data tapes produced by the base-line

inspection conducted with Intercontrole multifrequency equipment.

The DAARR and the two Analysis Round Robin experiments were designed to

provide data for estimating the variability in inspection results. The DAARR

showed the variability related to personnel differences for the overall

inspection (data acquisition and analysis) process using the same

instrumentation. Conversely, data from the two Analysis Round Robins provided

an estimate of the variability associated with only the analysis and

interpretation of recorded EC signals. The AATRR was designed to provide NDE

equipment developers with the opportunity to test equipment and/or analysis

procedures on service-degraded tubes in exchange for providing inspection

results. The intent of this round robin was not to compare the various

techniques, since some of the techniques were designed to detect specific

defect conditions, but to determine the improvement in defect detection and

3.38
SERIES 51

.(4 ID 00 0: C '1 q: N.M IN N Cq ml m• ('

-..----- 46 45 ---- 44 43 -- 42

41 ..--- 40 --- .- 39

37

35 ... 34 .. - 33

to

-,!7!-: 21

20

19

-i I6 -- 175

--- 14 1

-- - 6----5

-4W3

ROS

"--NOZZLE MAN WAY------*

FIGURE 3.21. Subset of Tubes Selected for Round Robin Inspections
sizing to be expected from emerging NDE technologies. Results of the round

robin exercises were described in a report by Doctor et al. (1988b).

3.4.4.1 Data Acquisition and Analysis Round Robin (DAARR)

The DAARR comprised five teams investigating the round robin subset of

steam generator tubes using the Zetec MIZ-12 multifrequency eddy-current data

acquisition system and a Zetec DDA-4 data analyzer. Four of the teams were

selected from proposals received in response to a request for proposal, and

one team was provided by a SGGP sponsor. All teams were provided by commercial

vendors that perform eddy-current NDE services. A minimum of four teams was

determined to be necessary to provide a statistically valid measure of NDE

reliability. Each team was provided with information on the operational

history of the steam generator, along with a work statement describing test

objectives, applicable codes and personnel qualification requirements. In

addition, several test parameters were standardized to eliminate extraneous

variables from the inspections. This included a protocol to control testing

procedures and reporting of results. Three of the examination frequencies

were specified: 400 kHz using the differential mode and 100 kHz using the

differential and absolute modes. A fourth optional frequency was selected by

each team at its own discretion. While the teams were required to take data

on the three designated frequencies, they were encouraged to analyze the

signals by their standard field practice procedures. Each team had five

working days to acquire and analyze the data from 320 tubes. Most teams

completed the task in less time.

A Zetec Type SFM EC probe was specified. The SFM type, typical of those

used for U.S. steam generator examinations, contains a permanent magnet which

is designed to saturate ferromagnetic elements in the tube (permeability

variations), canceling their affect on the EC signal. The teams were

instructed to use specific probe sizes in specific tubes. This was done to

ensure the maximum extent of inspection of each tube by using probes with the

largest fill factor that would pass through the tube. The probe sizes were

either 0.700 in. or 0.720 in. in diameter. All teams used the same probe

positioning equipment and probe pusher-puller, supplied by PNL.

Table 3.5 provides a summary of the origin and types of wall-loss

indications reported by each of the teams participating in the DAARR. Strictly

speaking, "indication" can mean the presence of an EC signal; however, EC

inspections of this generator produced many signals, some of them caused by

denting and conductive deposits, which are not directly related to tube

failures. Henceforth, "indication" and "wall-loss indication" are used

synonymously, both referring to a tube wall degradation. The first line of

each table shows the total number of indications reported by each team.

The second line of the table shows the number of indications after

multiple indications were removed. During the base-line examination, there

were numerous instances of multiple indications reported at the top of the tube

sheet. This noticeably complicated the task of comparing the two inspections,

since it was often difficult to ascertain which indications from the two

3.40
TABLE 3.5. Numbers and Types of Indications Reported by

Acquisition and-Analysis Round Robin Teams

Numbers of Indications

Team

A B C D E

Total Indications 287 207 292 267 270

Total Indications 255 205 226 265 223

(Multiples Removed)

OD Indications 234 195 226 261 212

ID Indi~cations 16 3 1

Unidentified Origin 5 7 3 11

Wastage* 2

Cracking* I

Possible IGA* 20

Unidentified Indica- 252 205 206 265 223

tion Type

*Reporting of type of indication not required.

inspections matched. Therefore,-the instructions for the round robin teams

explicitly stated that the location of the deepest part of an indication

should be reported; however, the decision of where one indication ended and

another began was a subjective judgmentof the data analyst. For the round

robins, the propensity *for reporting adjacent multiple indications varied

among the teams,. To avoid misrepresenting the numbers of common detections

caused by multiple indications, multiple OD calls within 3 in. on the same

tube were considered to reference different parts of the same large defect

and were combined. The resulting OD indication was given the size and

location of the largest (in terms of percent through-wall) of the original

multiple indications. Comparing the numbers in the first two lines of

Table 3.5 shows the effect of the removal of multiple indications.

The three lines of the tables show the number of indications in each

origin category, outer diameter (OD), inner diameter (ID), or unidentified

origin. Unidentified indication origins refer to reported indications where

no origin data was provided or the data provided was ambiguous. In most cases,

this data omission was probably inadvertent, although some teams seemed more

prone to leaving out information than others.

The last four lines of the table show the numbers of each type of

indication reported. The teams were not required to provide this information,

and some teams were somewhat more disposed to doing so than others. If no

indication type was provided, this was included in the unidentified type

category.

The data presented in Table 3.5 shows that the total number of wall-loss

indications ranged from 207 to 292. The removal of multiple indications

reduced the range of the number of indications reported from 205 to 265. The

number o- OD indications reported varied from 195 to 261. The number of ID

3.41
indications varied from 0 to 16. Most teams elected not to report indication

types, although Team C did report 20 tubes contained intergranular attack

(IGA). In all but two cases, these were associated with reported OD wastage

at the top of the tube sheet.

For comparison purposes, note that the number of indications reported from

the base-line inspections of the round robin tubes is 242 for Team X and 292

for Team Y.

The reference locations of the indications reported by each DAARR team are

given in Table 3.6. The teams were instructed to reference indications to

the tube sheet or nearest support plate. When this did not occur, the

indication locations were established during statistical analysis. In the

case where the nearest reference point was the seventh support plate, this

instruction apparently caused confusion. Some teams referenced indications

up to 14 feet into the U-bend to the seventh support plate. Other teams

referenced U-bend indications as simply "U-bend," thereby providing no means

to unambiguously locate the indication. The reference locations of indications

at the seventh support plate and U-bends have been left the way the teams

reported them. Some of the indications referenced to the seventh support

plate were, therefore, actually in the U-bend region. There were also cases

where no location was provided, an incomplete location was provided (for

example, "HL"), or the location reference was illegible. These were tabulated

in the "Unidentified Location" category.

Similar to what was observed for the base-line inspections (see

Table 3.3), a few wall-loss indications are scattered among the support plates,

with the majority of indications located at the top of the tube sheet on the

hot-leg side of the generator. The next highest number of indications reported

is usually at the cold-leg top of tube sheet. Again, it is important to

remember that similar numbers of reported indications by two or more teams

does not imply the teams are referring to the same defects. However, it is

interesting to compare the total numbers of indications reported by the teams

at given locations. Table 3.6 shows that the number of indications at the

hot-leg top of tube sheet was relatively constant throughout the five

Acquisition and Analysis Round Robin inspections, ranging from 192 to 223.

An area of significant disagreement in numbers of reported indications was

the cold-leg top of tube sheet.

3.4.4.2 Advanced/Alternate Techniques Round Robin (AATRR)

As previously indicated, this round robin consisted of nondestructive

inspection with equipment or analysis procedures different from those used in

the base-line or other round robin inspections discussed in the preceding

paragraphs. While it was encouraged for these examinations to be conducted

using the round robin subset of generator tubes, it was not required. In

fact, most of the advanced technique examinations utilized special equipment

which was designed to characterize a specific defect type or inspect a

specific region of the generator. Also, several of the techniques utilized

in this category were experimental or developmental in nature. While this

inhibited direct correlation of results with previous work, this effort was

3.42
TABLE 3.6. Location of Indications Reported by

Acquisition and Analysis Round Robin Teams

Numbers of Indications Location

Team

Reference A B C D I

Hot Leg

,Tube End

Tube Sheet Roll

Tube Sheet Crevice

Top of Tube Sheet

Support Plate 1

Support Plate 2

Support Plate 3

Support Plate 4

Support Plate 5

Support Plate 6

Support Plate 7

Cold Leg

Tube End

Tube Sheet Roll

Tube Sheet Crevice

Top of Tube Sheet

Support Plate I

Support Plate 2

Support Plate 3

Support Plate 4

Support Plate 5

Support Plate 6

Support Plate 7

U-bend

Unidentified Location

8

2

216

3

192 212 223 218

1

1

1 1

6 4

10

6

2

4

1

5

1.

1

2 4

8 36

1

1

1

1

1

1

intended to provide a validation opportunity for potentially improved

inspection techniques.

Examinations under this task, with the exception of the profilometry, were

conducted at participant expense. Eddy-current NDE, utilizing special eddycurrent probes, included tube sheet crevice examinations with an 8x1 pancake

probe and with a French rotating eddy-current probe. A special three

frequency double mix, using a MIZ-12 was unsuccessfully attempted by another

team to define tube damage under support plates. A developmental multifrequency/multiparameter eddy-current system was applied to the generator.

This examination experienced difficulties due to failure of special eddycurrent probes with surface-riding pancake coils. The ridges left during

plug removal caused particular problems for this system. In addition, the

system anlysis software required modification to allow a greater range of

3.43
signal inputs without saturation. A Japanese examination concentrated on

characterization of inner row U-bends for stress corrosion cracking, with a

special flexible probe, and at studying the tube-sheet crevice area for

primary-side stress corrosion cracking at the roll transition. A surfaceriding four-coil focused probe was used. One team utilized computer-aided

data interpretation. This involved an in-line computer system for signal

gating, mixing and data analysis (Doctor et al. 1988b).

Two teams applied ultrasonic (UT) inspection procedures. One team

utilized a rotating UT probe, the other an electronically sequenced probe

with eight sections. Both probes were utilized to inspect the tube sheet

region for intergranular attack (IGA) and stress corrosion cracking (SCC).

One team utilized UT to characterize defects found near the top of tube sheet

with eddy-currents. The UT sizing results were used to modify the eddy-current

reported results throughout the generator.

Table 3.7 shows the number of wall-loss indications reported during the

advanced/alternate techniques round robin. It is not appropriate to compare

all teams with each other (or with the other round robin teams) based on the

number and type of indications, since one team inspected only a subset of the

round robin tubes. The advanced/alternate techniques were often designed

specifically to detect certain types of defects that a conventional bobbincoil inspection would not detect.

TABLE 3.7. Numbers and Types of Indications Reported

by Advanced/Alternate Technique Teams

Numbers of Indications

Team

U* V UU VV

Total Indications 151 279 119 130

Total Indications 145 279 118 126

(Multiples Removed)

OD Indications 145 257

ID Indications 22

Unidentified Origin 118 126

Wastage** 48 256

Wastage/Cracking** 2

IGA* 95

Unidentified Indica- 22 118 126

tion Type

*Team U inspected a subset of round robin tubes.

**Reporting of type of indication not required.

3.44
The locations of advanced/alternate techniques wall-loss indications are

given in Table 3.8. As the advanced/alternate techniques inspected different

numbers of tubes (Team U inspected a subset of the round robin tubes) and the

other teams inspected different locations within the tubes, it is not

appropriate to compare the teams based on numbers of reported indications

given at any location other than the hot-leg top of the tube sheet.

TABLE 3.8.

Location

Reference

Location of Indications Reported by

Advanced/Alternate Technique Teams

Numbers of Indications

Team

V UU VV

Hot Leg

Tube End

Tube Sheet Roll

Tube Sheet Crevice

Top of Tube Sheet

Support Plate 1

Support Plate 2

Support Plate 3

Support Plate 4

Support Plate 5

Support Plate 6

Support Plate 7

Cold Leg

Tube End

Tube Sheet Roll

Tube Sheet Crevice

Top' of Tube Sheet

Support Plate 1

Support Plate 2

Support Plate 3

Support Plate 4

Support Plate 5

Support Plate 6

Support Plate 7

U-bend

Unidentified Location

56

.6

2

1

2

10

47

1

4

2

3

10

239

3

2

95 104

7

2

1

17 16

7

2

19

1

1(1)

2

3

I 1

*Team U inspected a subset of round robin tubes.

()Denotes indications referenced between support plate and

next higher support plate.

3.45
Under subcontract, Babcock and Wilcox performed a profilometry examination

using the PROFILE-360 system. The primary purpose was to provide a measure of

deformation from which strain levels in dented tubes are computed, for

comparison to metallographic analyses. The presence or absence of IGSCC as a

function of denting-induced strain is an issue in developing/evaluating a

strain-based tube-plugging criteria. Profilometry examinations were made in

100 tubes, including some round robin tubes, and a distribution of tubes

through various regions of the generator. Results of this examination are

detailed in a report by Doctor et al. (1988b) and summarized in Table 3.9.

TABLE 3.9. Statistical Summary of Profilometry Data

Number of Tubes Examined: 101

TSP NUMBER MAX DIA SIGMA MIN DIA SIGMA MAX STRAIN SIGMA

7 91 781. 18. 731. 40. 8.3 13.5

6 91 781. 12. 736. 20. 6.7 5.1

5 91 783. 12. 733. 19. 6.9 3.6

4 92 785. 17. 728. 29. 7.7 6.0

3 95 786. 17. 727. 39. 8.4 10.3

2 99 791. 24. 723. 36. 8.9 8.3

1 101 789. 21. 726. 32. 9.1 10.2

0 101 771. 10. 755. 4. 2.6 1.1

SUM 761 783. 18. 732. 31. 7.3 8.4

Distribution Summary

DIAMETER REDUCTION (MIL) STRAIN (%)

TSP > 20 > 50. > 100. > 10.0 > 15.0 > 25.0

7 39 12 4 8 2 4

6 51 10 1 7 2 2

5 55 14 1 11 1 1

4 61 12 5 7 3 2

3 58 12 5 8 4 3

2 58 16 9 11 6 5

1 52 15 8 13 5 8

0 0 1 0 1 0 0

SUM 374 92 33 66 23 25

3.4.4.3 Analysis Round Robin I

The eddy-current data tapes from the base-line examination conducted with

Zetec multifrequency equipment were copied for 424 tube scans. These data

were from scans of the 320 round robin tubes as conducted in the DAARR, plus

additional scans, including multiple scans of a tube with different probe

sizes. Each inspection was assigned a random identification number which was

recorded on the tape and used in place of the tube row and column numbers.

Six teams (analysts) were selected through a competitive procurement. In'

3.46
addition, 3 teams participated through Project sponsors. The teams were not

required to perform their data analysis at PNL and were given substantially

more time to perform the analysis than the DAARR teams had been given. They

were allowed to use any data analysis system; however, all teams chose to use

the Zetec DDA-4. Detailed reporting instructions were provided to each team

including a visit by a PNL eddy-current NDE analyst. The participants were

provided with historical information on the Surry steam generator similar to

that provided the DAARR teams. The 424 tube scans were codified so that data

on tube location was not available.

Table 3.10 contains data from Analysis Round Robin I (Doctor et al.

1988b). This round robin produced the largest variation in number of reported

indications of any of the round robins. The smallest number of indications

reported-was 186 while the largest was 630. Removal of multiple indications

did not reduce the range significantly: 179 to 600. Discounting Team K's

indications, which had missing information on the origin of the indications,

the number of OD indications ranged from 163 to 588. The number of ID

indications ranged from 0 to 156. While the acquisition and Analysis Round

Robin teams elected not to report the types of indications, four of the nine

analysis round robin teams (FF, J, I, and M) reported a substantial number.

Of the four, three (FF, J, and I) reported mostly wastage, while the fourth

team (M) reported mostly cracking.

TABLE 3.10. Numbers and Types of Indications Reported

by Analysis Round Robin I Teams

Numbers of Indications

Team

F FF G H I J K L M

Total Indications

Total Indications

(Multiples Removed)

OD Indications

ID Indications

Unidentified Origin

Wastage*

Cracking*

Possible IGA*

Unidentified

Indication Type

227 196 401 359 243 630 186 190 334

222 196 377 358 227 600 182 179 313

221

1

196

196

221 262 219 588

156 96 6 11

2 1

156 573

69 21

3

377 358 2 3

17 163 301

12

165 16

2

175 179

67

234

222 12

*Reporting of type of indication not required.

3.47
Table 3.11 shows the locations of wall-loss indications reported by

Analysis Round Robin I teams. The numbers of indications reported at the

hot-leg top of tube sheet differs by a maximum of 71 indications. This is

notable, in that the difference between the smallest and largest total number

of indications reported is 421. The areas of greatest disagreement in terms

of absolute numbers of indications were the hot-leg second support plate and

the cold-leg top of tube sheet.

TABLE 3.11. Location of Indications Reported

by Analysis Round Robin I Teams

Numbers of Indications Location

Location Team

Reference F FF G H I J. K L M

Hot Leg

Tube End

Tube Sheet Roll

Tube Sheet Crevice

Top of Tube Sheet

Support Plate 1

Support Plate 2

Support Plate 3

Support Plate 4

Support Plate 5

Support Plate 6

Support Plate 7

Cold Leg

Tube End

Tube Sheet Roll

Tube Sheet Crevice

Top of Tube Sheet

Support Plate 1

Support Plate 2

Support Plate 3

Support Plate 4

Support Plate 5

Support Plate 6

Support Plate 7

U-bend

Unidentified Location

1

4

208 159 206

3

2 3

1

3

7

1

1 10

137(8)

1(10)

5(3)

5(8)

2(12)1*

0(3)

2(5)

1

192 208

2 5

99

61

10

15

14

1 112

163 169 147

6

1 5

6

1

7

5

1 1 79

3

10

1

37 107

7

3

3

4

1

2

14

109(1)1*

0(6)

1(2)

0(6)

3(4)

1(3)1*

3(2)

2

28 37

2

3

6

2

3

9

4 11

14 9 32

1

1

3

1

1

9

9 1

2 1 2 3

referenced between support plate and next higher

referenced between support plate and 2nd next higher

1

( Denotes indications

support plate.

* Denotes indications

support plate.

3.48
3.4.4.4 Analysis Round Robin II

This round robin was similar to the Analysis Round Robin I, only utilizing

data copied from the base-line eddy-current tapes produced with Intercontrole

multifrequency equipment (Doctor et al. 1988b). Six European teams who

routinely use this type equipment (an Intercontrole IC4AN analyzer)

participated under the direction of the French SGGP sponsor. Slight

differences to the other analysis round robin existed. Some of the tube

signals examined were different and the data tapes included actual row and

column information.

Table 3.12 shows the numbers of reported wall-loss indications from

Analysis Round Robin II. The variation in numbers reported was much less in

this round robin than for Analysis Round Robin I. The smallest number of

indications reported was 254 while the largest number reported was 376. The

range with the multiple indications removed was 227 to 291. The ranges of

numbers of ID and OD indications were also relatively small, despite the

substantial amount of origin information omitted by Teams 0 and Q.

TABLE 3.12. Numbers and Types of Indications Reported

by Analysis Round Robin II Teams

Numbers of Indications

Team

N 0 P Q R S

Total Indications 254 316 337 284 340 376

Total Indications 227 283 270 272 258 291

(Multiples Removed)

OD Indications 212 235 265 200 252 266

ID Indications 8 6 4 6 5 7

Unidentified Origin 7 42 1 66 1 18

Wastage* 102

Cracking* 11

Pitting 1

Shock* 1

Unidentified 227 283 155 272 258 291

Indication Type

*Reporting of type of indication not required.

For Analysis Round Robin II teams, the largest between-team difference in

number of indications was reported at hot-leg top of tube sheet (Table 3.13).

The second area of least agreement in terms of numbers of reported indications

was consistent, with the other two round robins, the cold-leg top of tube sheet.

3.49
TABLE 3.13 Location of Indications Reported

by Analysis Round Robin II Teams

Numbers of Indications

Location Team

Reference N 0 P Q R

Hot Leg

Tube End

Tube Sheet Roll

Tube Sheet Crevice I

Top of Tube Sheet 211 246 225 211 224 243

Support Plate 1 (1) 1

Support Plate 2 4 1 4 4 2(1)

Support Plate 3 1 3 2(1) 2

Support Plate 4 (1) 1

Support Plate 5 3 1(2) 2.

Support Plate 6 1 3

Support Plate 7 5 1 1 2 2 7

Cold Leg

Tube End

rube Sheet Roll

Tube Sheet Crevice

Top of Tube Sheet 18 26 47 15 21

Support Plate 1 1 1 1 1 1

Support Plate 2 1 1

Support Plate 3 1 1 10() 1

Support Plate 4 1

Support Plate 5 1 1

Support Plate 6 1

Support Plate 7 1 3 2 1 3

U-bend 8 3 3 1

Unidentified Location 2 3 2 6

()Denotes indications referenced between support plate and

next higher support plate.

3.5 SPECIMEN REMOVALS AND NDE VALIDATION

A prime purpose in utilizing a removed-from-service steam generator was

to accurately characterize service-induced degradation, mainly to provide

correlation with the results of the nondestructive inspections. Thus, the

removal of specimens from the generator, without altering their

characteristics, was of prime importance. This was also a formidable task,

considering both the physical magnitude of the operation and the fact that

the work involved sectioning radioactive material.

3.50
To validate the NDE round robin results and.determine the condition of the

tubing, tube segments were-selected and removed from the generator for visual

and metallurgical examination. This section describes the logic used for

selecting specimens from the.hundreds of. tubes..inspected during the various

round robins, the methods used for removal of these specimens, the results of

the visual and metallurgical examinations, and results of burst tests performed

on selected specimens. A more detai.led descriptionhas been presented by

Bradley et al. (1988). d i has b p t

3.5.1 Specimen Selection

The purpose of the specimen removal plan was to select tube sections for

removal, and visual and metallographic examination toensure that a

technically valid estimate of the reliability of the NDE of steam generator

tubes could be made. A detailed description of the specimen removal logic,

methods and validation results is presented in the Task 13 report (Bradley

et al. 1988). The foilowing criteria had to be met:

o Sufficient numbers of specimens with.defects were needed to estimate the

probability of detection (POD) of the inspection methods as a function

of defect size.

" A representative cross secti'onand a sufficient number of defects were

required to estimate the POD for different types of defects in the

generator.

* A representative selection and number of specimens without reported

defects were-neededto estimate the "false negative" (i.e., no defect

reported when one ex-ists) error rate.. .

The selection of the tube specimens for the NDE validation was governed

by the following considerations:

" The total number of tube sections that could be removed based on the

cost of post-removal examination and analysis was 500-to 600.

* With a few exceptions, the selected tube section had to have been

inspected by all of the DAARR teams to ensure that the maximum

information was obtained for NDE reliability.

Since the purpose of the validation was to establish the reliability of

NDE, there was a need to have a large number of specimens with defects from

which to estimate the POD. Three major strata (categories) of potential

specimens were defined to reflect the lack of direct knowledge of the physical

condition of the tube sections and to ensure that proper emphasis was given

to the selection of a large number of specimens with defects. The three strata

were as follows:

o tube sections with a possible defect based on evidence of a reported

defect indication by at least one inspection team

3.51
* tube sections with a potential defect based on past history but with no

indication reported

" tube sections in which a defect was not expected and for which no

indication was reported.

The number of specimens to be allocated to each stratum was based on the

goal that 80% of the specimens should contain defects. It was felt that

slightly more weight should be placed on the specimens with reported

indications; thus, Stratum 1 was to contain about 50% of the specimens;

Stratum 2, 30%; and Stratum 3, 20%.

Each stratum was then subdivided into finer substrata to ensure proper

representation of the various locations within the generator.

Table 3.14 summarizes the numbers of specimens in each of the strata by

the major regions of the generator (tube sheet, support plate, U-bend). The

first line for each stratum lists the total number of specimens. The

subsequent lines give the number of specimens in the substrata. Of the 556

specimens removed from the generator, 48% were from Stratum 1, 30% from

Stratum 2, and 22% from Stratum 3. This demonstrates that the specimen removal

plan was satisfied. The division of specimens into regions of the generator

TABLE 3.14. Numbers of Specimens per Strata and Major

Location Within the Generator

No. of Specimens in Region

Stratum Tube Sheet Support Plate U-Bend

1. Reported 191 54 24

indications

Defects 159 27 21

Advanced technique 32 27 3

defects

2. Defects expected 65 71 29

Sludge pile 34

Crevice 19

Roll transition 12

Strain 29

Hard/soft spots 42

Inner row U-bends 29

3. Defects not expected 16 78 28

Conditions 16 39 1

Straight sections 39

Outer row U-bends 27

3.52
is as follows: 49% in the tube sheet, 37% in the support plates, and 14% in

the U-bend region. Although the tube sheet was the location of most of the

reported indications, it accounted for only half of the specimens; the

support-plate and U-bend regions of the generator were adequately represented.

3.5.2 Specimen Removal

Removal of specimen tubes from the steam generator was a critical step

in the NDE validation effort. Proper specimen identification and removal with

minimal damage to the tube sections was essential for comparing postremoval

inspection results with data from in situ NDE. The large number of specimens

and widespread distribution within the generator required the development of

complex methods and procedures for identifying and removing specimens from

each region of the generator. The procedures and methods used for removing

tubes from the U-bend, tube support-plate, and tube sheet regions are

discussed in the following sections (for more detail see Bradley et al. 1988).

3.5.2.1 U-Bend Region

About 2600 U-bends were removed from the generator during the U-bend

removal operation. U-bend specimen tubes were identified by inserting a probe

in an adjacent cut tube from the channel head, where identification of tube

locations could be made. The specimen was then marked with a dry paint marker

on a pole.

A pair of abrasive cut-off saws were the main tools used to remove U-bend

specimens. A picture of one of the cut-off saws is shown in Figure 3.22.

The two saws were located opposite each other and at 900 to the tube lane.

Shell penetrations just above the seventh TSP permitted cut-off wheel access

to the tube bundle. These tools were operated semiremotely as shown in

Figure 3.23. U-bends were cut flush with the top of the seventh TSP from

both the hot- and cold-leg sides. The U-bends were subsequently lifted out

through the top of the generator. To permit handling within the confines of

the SGEF, the larger outer row U-bends were cut in half with a hydraulically

assisted shear, prior to removal.

3.5.2.2 Tube Support-Plate Region

One of two methods was used to remove specimens from the support-plate

region of the steam generator, depending on specimen location. Specimens that

included the seventh TSP intersection or that contained defect indications

<1 in. above the seventh TSP were removed as individual specimens concurrently

with the U-Bend specimens. The remaining specimens from the support-plate

region were removed by full-length tube pulls following the U-bend removals.

Specimens that included the seventh TSP were removed concurrently with or

immediately after removal of the U-bends. The specimen and adjacent tubes

were cut about 6 inches above the seventh TSP with the abrasive saw, and the

U-bend portion of the tube was removed. The sample tube was then positively

identified by inserting a probe from the channel head. As the probe emerged

from the cut tube, the specimen was marked with a dry paint marker. Cutting

the speciiren off below the seventh TSP required an access hole in front of

3.53
FIGURE 3.22. Abrasive Cut-Off Wheel

FIGURE 3.23. Semi-Remote Abrasive Wheel Cutting at Top of Seventh TSP

3.54
the specimen for a plasma arc torch. Thus, several adjacent tubes were cut

from above with an internal tube cutter and a hole was broken in the seventh

TSP. The specimen was then cut off below the seventh TSP with a plasma arc

torch and removed from the generator.

Removing specimens from below the seventh TSP was an especially difficult

problem, and several approaches were considered. An early plan to penetrate

the shell at many locations and to tunnel in from the side to retrieve

specimens was discarded because of the following serious disadvantages:

1) the diverse location of the specimens within the tube bundle would have

required a large number of shell penetrations; 2) unsupported portions of

tube bundle might collapse or otherwise require support; and 3) the potential

for incorrectly identified specimens was deemed unacceptably high.

Consequently, specimen removal by pulling tubes from the top of the tube bundle

was considered the most practical approach because it eliminated or minimized

these difficulties.

A major advantage in pulling tubes from the top of the bundle was the

ability to positively identify specimen tubes at the seventh TSP. A Zetec

template plug was tagged with the row, column, and hot-leg/cold-leg information

and inserted into the probed tube. After all specimen tubes had been tagged,

photographs were taken to provide permanent records of the tube

identifications.

Full-length tubes were pulled by first cutting the tube with an internal

tube cutter at a point above the TTS. Once cut, the tube was then gripped at

an appropriate point and pulled out with the SGEF bridge crane as shown in

Figure 3.24. Tubes removed by this method were minimally damaged because the

TSPs broke apart during the pulling operation. On the hot-leg side of the

generator, all seven TSPs were so heavily corroded that they broke easily.

Damage of the specimen tube was kept to a minimum by first pulling several

adjacent non-specimen tubes, which broke the surrounding TSPs and freed the

tube of interest. Generally, TSPs 1 through 4 on the cold-leg side would not

break. Therefore, cold-leg specimens with defect indications above the fourth

TSP were pulled after pulling neighboring tubes, until the specimen tube could

be gripped below the defect zone. Cold-leg specimens with defects below the

fourth TSP were also pulled after surrounding tubes had been removed to reduce

the force necessary to pull the specimen tube. In no case was the applied

force great enough to break the tube, but several were plastically deformed

during the process.

3.5.2.3 Tube Sheet Region

Most specimens removed from the tube sheet and sludge pile region of the

generator were retrieved by jacking the tubes out of the tube sheet into the

channel head. A multi-step process was used: 1) internally cutting the tube

above the sludge pile, 2) inspecting the cut tube using a video probe,

3) removing the tube-to-tube-sheet fillet weld, 4) heating the roll-expanded region to free the tube, 5) inserting the pulling mandrel, and 6) attaching

the hydraulic jack and pulling the tube. Each tube specimen to be removed

was cut above the sludge pile from the inside with a rotary hand cutter. A

completelj cut tube was essential for reducing the pulling forces and

3.55
4

FIGURE 3.24. Full-Length Tube Being Pulled Up by SGEF Bridge Crane

minimizing tube damage, so the cut tube was inspected with a video probe

before removal. Next, the tube-to-tube-sheet fillet weld was milled off. A

fixture was used to hold the air motor and cutting tool in proper alignment

against the tube sheet. The fillet weld was then removed by raising the motor

and single-blade cutting tool using the lever device of the holding fixture.

After the fillet weld had been removed, the roll-expanded region was heated

with an oxyacetylene torch, which reduced the force required to pull the tube

from the tube sheet. A threaded mandrel was then inserted into the tube end.

A 2-ton hydraulic jack was positioned over the mandrel and the specimen was

pulled out as the jack operated on the mandrel. A photograph of a tube being

removed from the tube sheet is presented in Figure 4.25. The mandrel was

removed from the tube by grinding an axial slit along the lower three inches

of the tube end.

3.56
FIGURE 3.25. Tube Pulling Through Tube Sheet;

Hydraulic Jack in Operation

Some specimens that had been plugged during service, in which the plugs

had subsequently been removed by drilling (Wheeler et al. 1984), fractured

during the tube-pulling operation. The fractures usually occurred in the rollexpanded region, although some failures occurred above the roll transition

but within the region drilled out to remove the plugs. The fractures were

caused by weakening of the tube due to the presence of IGSCC. The IGSCC

probably formed because of incomplete bonding, leading to a crevice

environment and high residual stresses due to plug installation. These

specimen tubes were removed by repetitive insertion and jacking of mandrels

with various diameters and lengths until the complete specimen was removed.

In addition to the specimens removed by tube pulling, two hot-leg and one

cold-leg sections of tube sheet, each containing nine tubes in a 3x3 array,

were removed from the generator. The tube sheet sections were successfully

removed by an overbore drilling technique after initial attempts using metal

disintegration machining were unsuccessful. The process was developed by

3.57
FIGURE 3.26. Tube She(

Channel H

Framatome, Inc., of France;

illustrated in Figure 3.26.

holes surrounding the desire

The apparatus was essentiall

te

ie,

Section

ad

Drilling Equipment in Place in

the removal technique and equipment are

The process consisted of over-boring the tube

d section so that the holes overlapped slightly.

y a modified lathe operating in a vertical

3.58
position. Two oil holes running the full-length of each drill bit provided

lubrication to the cutting edges. Approximately one inch before breakthrough,

oil flow was stopped to prevent contamination of the crevices. After all the

surrounding holes were drilled, the tubes in the section were internally cut

above the sludge pile to free the section for removal. Sixteen tubes were

overbored to release one 3x3 tube sheet section.

3.5.3 Validation of NDE Round Robins

The purpose of the NDE validation studies was to determine the reliability

of NDE, primarily bobbin-coil eddy-current, to detect and size defects in steam

generator tubing. A secondary purpose was to establish the condition of the

tubing throughout the generator and identify degradation mechanisms in regions

normally inaccessible to direct visual or destructive examinations. Tube

segments removed from the generator were prepared for validation by

1) reducing the length to minimize radiation exposure to personnel and

facilitate handling, 2) permanently identifying each specimen by inscribing a

unique identification number on the tube surface, 3) deburring the cut ends

for safety, and 4) chemically removing the deposits to enable visual inspection

of the tube surfaces. Post-removal eddy-current inspections were conducted

on specimens both before and after the deposit removal. These inspections

aided the validation by establishing the effect of the deposits on EC signal

characteristics.

Detailed visual examinations were a major facet to the specimen validation

effort. The outside surfaces of all specimens and the inside surfaces of

selected specimens were examined in detail using a stereo microscope at

magnifications ranging from lOX to 70X. The type and location of tube

degradation observed was recorded and a visual estimate of wall-loss for

surface defects such as pitting and wastage was made. In general, it was not

possible to estimate the depth of cracking or intergranular attack (IGA) by

visual means. Three ranges of wall-loss were used in making the visual

estimates: light (<20%), medium (20 to 40%), and heavy (>40%). These

estimates provided a qualitative ranking of specimen degradation and were

used in selecting specimens for destructive metallographic examinations and

burst testing.

In selecting specimens for destructive metallographic examination,

emphasis was given to validating defect indications reported by NDE, which were

primarily located at the hot-leg TTS. Selections were made to obtain a range

of defect distributions and depths within the metallographic data in order to

establish the accuracy of NDE in detecting and sizing tube defects.

Metallographic sections were also prepared for a limited number of specimens without significant visible degradation to ensure that unidentified defects

were not present or to evaluate the nature of the OD deposits. The type of

degradation observed and the maximum wall-loss were recorded for each specimen.

The visual and metallographic data were assembled in a computer data file

that could be combined with the EC inspection results from the base-line and

round robin experiments. Statistical evaluation and analysis of the combined

data provided estimates of NDE reliability in detecting and accuracy of sizing

3.59
steam generator tube defects. A detailed report of the NDE validation studies

has been prepared by Bradley et al. (1988).

3.5.3.1 Metallurgical Validation Results

Metallurgical examinations of the validation specimens included detailed

visual inspection of all specimens and metallographic examination of selected

specimens. The visual examinations identified the type and location of tube

degradation and provided a qualitative estimate of wall-loss for surface type

defects, such as wastage, pitting, etc., while a more precise estimate of

wall-loss and defect characterization was obtained from the metallography.

The following types of defects and conditions were found during the

metallurgical examinations:

* U-Bend Region

- anti-vibration bar (AVB) contact wear with corrosion

- stress corrosion cracking (SCC) in Row 1 and 2 U-bends

- grinder marks

- Cu-rich surface deposits

* Support-Plate Region

- within TSP intersections

* denting

" pitting, wastage, and IGA

" SCC

" Cu-rich surface deposits

- between TSP intersections

" pitting and wastage

" hole

" Tube Sheet Region

- sludge pile above TTS

" pitting

" wastage

* SCC

- roll expansion

e SCC in plugged tubes

A brief description of these defects is given below.

3.5.3.1.1 U-Bend Region. Shallow wear and corrosion at AVB contact

areas, SCC at the apex of Row 1 and 2 U-bends, and two grinder marks produced

during an inservice repair operation were the only defects found in specimens

from the U-bend region of the generator. Evidence of wear and/or corrosion

was found on 13 of the 70 AVB contact areas examined, see Figure 3.27.

Metallographic sections through seven of the most severely-degraded areas

showed a naximum wall-loss of 25% for one specimen. The wall-loss for the

3.60
p I

50 pm

FIGURE 3.27. Wear and Corrosion at AVB Contact Areas on

U-Bend Specimens 416A and 416B (R31 C58)

remaining specimens was less than 12%. The SCC was found only at the apex of

Row 1 and 2 U-bends. Both OD and ID initiated cracks were observed in Row 1

U-bends, while only OD initiated cracks were found in Row 2 specimens.

Figure 3.28 provides an example of the cracking observed in a Row 2 U-bend.

The IGSC cracking was attributed to high stresses produced from the inward

movement of the seventh TSP into the flow slots due to denting. SCC was not

observed at the apex of Row 3 U-bends or at the bend transition region of any

specimen examined.

For the defects found during validation, the two grinder marks were the

only U-bend defects detected by the EC inspection teams. However, Row 1 and

2 U-bends were not inspected because of severe denting at support plates and

tube ovalization at the U-bend apex. The lack of detection of the AVB contact

wear is likely associated with the shallow nature of the defects and possible

interference from the AVB and copper-containing deposits on the OD surface

surrounding the AVB wear location.

3.61
I, A.

__ 200/pm __ __ 100 Am'

FIGURE 3.28. Optical Micrographs Showing Axial IGSCC at the Apex of U-Bend Specimen 1049 (R2 C27)
Tubes removed from the U-bend and support-plate regions of the generator

were covered with a dark granular deposit. The outer layer could be easily

removed by washing in water with a soft-bristled brush, leaving a tenacious

copper-colored deposit. This deposit generally covered the tube surface except

for small spots or axial strips where the Inconel was visible. A photograph

showing two spots and an axial strip where the metal is visible is presented

in Figure 3.29. Axial strips without the copper-colored deposits were often

found intermittently along the entire tube surface.

Metallographic examination showed that the thickness of the deposits

varied around the circumference of the'tube. Deposit thicknesses ranged up to

10 mils for the tubes examined. The deposits had a layered microstructure

with metallic Cu particles observed throughout. As described below, these

Cu-rich surface deposits were found to be the primary cause of false calls

made by the various inspection teams. Post-removal EC inspection indicated

that interruptions in the Cu-rich deposit gave rise to indications similar to

defects. Chemical cleaning of specimens to remove the Cu-rich deposit

followed by reinspection revealed that EC indications had disappeared.

FIGURE 3.29. U-Bend Specimen 450A (R14 C52 HL) Showing Discontinuities

in the Copper-Rich OD Surface Deposit

3.63
3.5.3.1.2 Support-Plate Region. Almost all of the tube degradation

observed in specimens from the support-plate region of the generator was

associated with the dented TSP intersections. Corrosion in the form of shallow

pitting, wastage, and IGA occurred'in the crevice region between the tube and

support plate in most of the 117 TSP intersections examined. Figure 3.30

shows a photograph of a typical TSP intersection. The severity of the OD

corrosion attack appeared greater in the more heavily dented tubes although

the visual estimate of wall-loss was generally less than 20% for all specimens.

Metallographic examination of 10 specimens with calculated strain values

ranging up to 39% confirmed the shallow nature of the OD corrosion as evidenced

by the data in Table 3.15. Various combinations of pitting (P), wastage (W),

and IGA were observed in the metallographic cross sections with the maximum

wall-loss ranging from 2 to 12%. Axial cracks were also visually observed on

the OD surface at regions of high tensile hoop-stress in specimens with

calculated strains greater than 60%.

Intergranular stress corrosion cracking that initiated at the inside

surface was also found in the more heavily dented specimens. The internal

surfaces of 27 hot-leg and 11 cold-leg specimens were examined, and axial SCC

was observed in 10 of the hot-leg specimens. An example of typical ID cracking

observed is shown in Figure 3.31. Maximum crack depth for the 10 specimens

ranged from 27% to 88% through-wall penetration. Calculated strain values

were available for 17 of the hot-leg specimens and provided a means for

evaluating the effect of tube deformation on the propensity for ID cracking.

Table 3.16 summarizes the results of visual and metallographic examination of

these specimens. The results may be divided into three strain regions: 0%

to 10%, 10% to 20%, and greater than 20%. No ID cracking was observed below

10% strain, while approximately one-third of the specimens with 10% to 20%

exhibited ID cracks. Above 20% strain, all of the specimens contained cracks.

Thus, strain-based plugging criteria between 17%-25% may not be low enough to

ensure a low probability for crack initiation. A definite relationship between

crack depth and calculated strain was not observed, which is consistent with

the complex stress distributions produced by the nonuniform nature of the

denting and temperature variations within the hot-leg-region of the generator.

Tube defects observed between TSPs consisted of a through-wall hole

produced during an inservice repair operation and-pitting/wastage type defects

immediately above the TSP in two specimens. Most of the EC inspection teams

identified the hole as a defect but, with few exceptions, did not locate the

two pitting/wastage type defects or any of the defects within the TSP

intersections. Apparently denting interfered with the EC signal and prevented

the defects from being detected. Only three of the specimens with ID cracking

were inspected by EC and no defects were reported from these inspections.

3.64
4M

001

Ln

FP- NOW -

5mm

FIGURE 3.30. Typical TSP Intersection; Specimen 665 (R7 C57 HL)
TABLE 3.15. Summary of Metallography Results Regarding

OD Corrosion of Dented TSP Intersections

Specimen No.

851

856

1041

921

909

925

935

926

947

455

Tube No.

R5 C26 HL1

R5 C77 HLU

R4 C51 HL1

R42 C47 HL4

RIO C39 HL5

R4 C51 HL5

R42 C52 HL5

R4 C51 HL6

R45 C55 HL6

R9 C47 HL7

Defect Type

Maximum

Wall-Loss,

Calculated

Strain, %

P/W

W/IGA

P/W/IGA

W

IGA

P/W/IGA

P/W/IGA

P/W/IGA

W/IGA

P/W

4

<2

12

<2

6

5

10

6

5

8

26.9

11.1

11.9

27

17

39.1

16.5

7

P = Pitting

W = Wastage

IGA = Intergranular Attack

5 mm

FIGURE 3.31. First TSP Specimen 1041 (R4 C51 HL) Showing ID Cracking

(26.9% strain)

3.66
TABLE 3.16. Summary

Interse

TSP

ID No. Row Col Location

of ID Cracking at Dented TSP

ctions With Various Strain Levels

553

567

525

455

922

929

921

909

927

923

924

947

935

908

1041

925

926

9

21

12

9

4

12

42

10

5

4

4

45

45

10

4

4

4

70

71

36

47

20

36

47

39

37

36

36

55

52

39

51

51

51

HL7

HL7

HL7

HL7

HL5

HL5

HL4

HL5

HL5

HL5

HL4

HL6

HL5

HL4

HLI

HL5

HL6

Strain, %(a)

4.2

4.9

5.4

7.0

9.9

10.1

11.1

11.9

12.5

13.4

14.1

16.5

17.1

17.9

26.9

27

39.1

ID(b)

Cracking

No

No

No

No

No

No

Yes

Yes

No

No

No

No

Yes

No

Yes

Yes

Yes

Crack(c)

Depth,

0

44

70

0

64

88

67

57

(a)

(b)

(c)

Based on profilometry data.

Based on visual examination.

Metallography results.

3.5.3.1.3 Tube Sheet Region. Several types of tube defects were found

in the sludge pile region immediately above the TTS.... These defect types

include pitting, wastage, localized circumferential corrosion (LCC), shallow

IGA, and OD initiated SCC. Pitting and wastage were the predominant defects

as can be seen in Table 3.17, which summarizes the visual inspection results.

Only 14 of the 241 specimens examined showed no evidence of corrosion in the

sludge pile region. Areas with wastage were observed on all the remaining

specimens, while additional areas with pitting and other forms of nonuniform

corrosion were found on 138 specimens. Cracking was found within the corroded

region of two hot-leg specimens. The severity of the degradation was much

greater on the hot-leg side of the generator, as evidenced by the differences

in the visual estimates of wall-loss.

Tube degradation within the cold-leg sludge pile consisted mostly of

wastage. Typically, tubes exhibited a slightly roughened surface and light

wall-loss, with small pits or pit clusters in the wasted region of some

specimens. Metallographic examination of 11 cold-leg specimens showed the

maximum wall-loss to range from 2% to 24%, with the average maximum wall-loss

being 10.7%.

3.67
TABLE 3.17. Summary of OD Visual Examinations of TTS Specimens

From the Sludge Pile Region of the Generator

Hot-Leg Specimens Cold-Leg Specimens

Degradation Number of Estimated Wall Number of Estimated Wall

Type Specimens Loss, % Specimens Loss, %

<20 20-40 >40 <20 20-40 >40

None 7 7

Pitting/ 105 10 46 49 23 22 1

Wastage

Wastage 15 10 5 0 74 73 1

Other(a) 9 3 3 3 1 1

(a) Includes various combinations of wastage, pitting, localized

circumferential corrosion and cracking.

The hot-leg TTS specimens also showed regions with wastage in the sludge

pile region for all of the degraded specimens. Wall-loss by wastage was

normally quite shallow with the estimated depth being less than 20%. However,

a few hot-leg specimens exhibited more severe wastage in the form of a ring

or groove that encircled the tube at various shallow angles. The axial extent

of this form of wastage was less than one-half inch, with a gradual taper in

wall thickness from both axial directions. Maximum wall-loss from this

ring-type wastage measured from metallographic sections ranged from 20% to

52%. An example of this type of degradation is given in Figure 3.32.

Although wastage was found in all degraded hot-leg TTS specimens, the

primary mode of degradation in most of the severely-degraded tubes was from

pitting and/or other localized corrosion processes. The pits were distributed

intermittently but not uniformly within the degraded region of the tube.

The axial extent of the degraded region varied from one specimen to another

but was typically less than 2 inches above the TTS. There was also a tendency

for clustering, which resulted in localized areas with a wide range of wallloss within the overall degraded region. The appearance of these more

severely-degraded areas is best described as clusters of overlapping pits,

although the mechanism by which they form may have included wastage or other

forms of corrosion.

Large variations in the number, severity, and distribution of degraded

areas (defect clusters) were observed in the hot-leg TTS specimens. A few

specimens exhibited a single pit or cluster of pits along with shallow wastage

in the corroded region. However, the majority of the hot-leg specimens

exhibited multiple-degraded areas that were often connected by shallow wastage.

The degra-ed areas were distributed both axially and circumferentially within

3.68
•2 m

FIGURE 3.32. Optical Micrograph Showing Ring Type Wastage on

Hot-Leg TTS Specimen 632 (R13 C37 HL)

the corroded region of the tube surface. Figures 3.33 and 3.34 show examples

of the variation in degradation observed around the tube circumference in

specimens with combined pitting and wastage. A few degraded areas were seen

in some sections, Figure 3.33, while others'showed extensive local areas of

degradation around most of the tube circumference, Figure 3.34. Maximum wallloss in specimens with pitting/wastage type degradation ranged up to 87%.

Most of the reported EC defect indications were located at the TTS and the

best agreement between inspection teams.occurred at the hot-leg TTS. The

severely-degraded specimens above the hot-leg TTS were generally reported by

most of the inspection teams, while detection was quite low for the lightly

degraded specimens from both the hot-leg and cold-leg regions of the generator.

Both hot-leg specimens having SCC within the corroded region of the specimen

were reported as defective by most of the inspection teams, see Figure 3.35.

However, it was not possible to determine if detection was based on the

presence of the cracks or the pitting/wastage typedegradation.

3.5.3.2 Evaluation of NDE Reliability

The reliability of the inservice inspection of steam generators depends

on the ability of the NDE technique to detect and size defects in the tubing.

To evaluate the probability of detection (POD) and sizing accuracy of EC

inspections, the results of the metallurgical validation data were combined

with the EC results from twelve inspection teams, two Base-Line inspections,

five Data Acquisition and Analysis Round Robin (DAARR) inspections, and five

,3.69
FIGURE 3.33. TfS Specimen

2 mm

FIGURE 3.34. Pitting/Wastage'Type Degradation, of a Hot Leg TTS Specimen

3.70.
I*.-%%"

100 PM 200 Mm

FIGURE 3.35. Optical Micrographs from a Longitudinal Section Showing OD Circumferential

Cracking of TTS Specimen 633C (R13 C44 HL at TTS + 0.2 in.)
Advanced/Alternate Techniques Round Robin (AATRR) inspections. The Base-Line

and DAARR teams inspected most of the validation specimens using bobbin-coil

technology, and their results are directly comparable. Except for one AATRR

team, limited inspection data were available from these teams, and comparisons

are therefore uncertain.

Table 3.18 gives a summary of the numbers of detections, false calls, and

nondetections from the visual examination data for each of the five DAARR

teams (A through E), the two base-line inspection teams (X,Y), and the five

AATRR teams (U, UU, V, VV, W). These calls are listed separately for the

three regions of the generator (TTS, TSP, U-bend). The detections and

nondetections for the OD defects are listed by the amount of wall-loss based

on a visual examination (L = light, M = medium, and H = heavy). Since the

depth cracks cannot be determined visually, the ID cracks are listed separately

as K. However, the two OD cracks that were found are included in the heavy

category because they were associated with pitting and wastage.

For the base-line and DAARR teams, the number of specimens inspected and

evaluated ranged between 481 and 497. The number of specimens inspected by

the AATRR teams ranged from 22 for Team W to 392 for Team V. For the DAARR

and the base-line teams, the numbers of wall-loss defects ranged from 312 to

324. The range was from 20 to 245 for the AATRR teams. Because of the small

number of specimens examined by the AATRR (with the exception of Team V), a

direct comparison with the base-line and DAARR teams is not appropriate.

For the TSP and U-bend regions, there may be more than one number reported

for a team, with the second number in parentheses. For the TSP region, the

first number is the number of reported and/or actual defects at the

support-plate intersection, which is the primary location for wall-loss

defects to occur. For the U-bend region, the first number is the number of

reported and/or actual defects at the AVB contact points. The number in

parentheses for both regions is the number of reported and/or actual defects

at other locations within a specimen.

Several patterns emerge from a study of Table 3.18. There is remarkable

consistency among the two base-line inspection and five DAARR (multifrequency

EC/bobbin-coil equipment) teams for the three areas of the generator. At the

TTS, where most of the defects occurred, the largest number of detections

occurred in the heavy size category and the largest number of nondetections

were in the light category. There was only one false call reported by one

team. There were 13 heavy wall-loss defects that were missed by at least one

of the teams. Seven of these were single team misses; five specimens were

missed by two teams, and one was missed by four teams.

In the support-plate region there were numerous nondetections of light

pitting and wastage at the support-plate intersections. The NDE inspections

were not able to pick them up because of the signal distortion caused by the

dented tube at these intersections. The one nondetection of a heavy wallloss defect was an OD pit that was not at the support-plate intersection, but

was missed by all DAARR and base-line teams. In addition, there were several

cracks in dented tubes at support-plate intersections that were missed by all

teams tha'. inspected that section of tube.

3.72
TABLE 3.18. Summary.Classification of Individual Team NDE

Inspection Results by-Region of the Generator

Data AcQuisition and Analvsis Round Robin

Type of

Defect(a) A

484

B

488,

Team

C

494

D

Number of specimens 487

Number of wall-loss

defects

E

481

312 3-17 324 -316 314

TTS

Detections

False Calls

Nondetections

TSP

Detections

False Calls

Nondetections:

U-bend

Detections

L

M

H

K

L

M

H

K

L

M

H

K

L

M

H

K

L

M

H

K

L

M

H

K

10

38

49

0

4

31

45

0

0 0 1 0 1 0

110

17

"' 2

0

116

25

6

0

107

15

3

0

94

13

1

0

110

14

1

0

13

41

48

0

26

43

50

0

1

0

0(1)

0

0

0

0(1)

0

0

0

0(0)

0

0

0

0(0)

0

0

0

0(i)

0

9

42

50

0

0(4) 0(2) 1(1) 0(1) 0

65(1)

1

0(0)

1

0

0

0(2)

0

69(1)

1

0(0)

2

0

0

0(2)

0

75'(1)

1

0(1)

3

0-1.

0

0(1)

0

68(1)

1

0(0)

2

0

0

0(2)

0

67(1)

1

0(1)

2

0

0

0(2)

0

False Calls

Nondetections

0 0 0. 0 0

12

1

0

0

12 12

1 1

0 00()

0 0

12

1

0

0

12

1

0

0

(a) L = Light, M = Medium, H = Heavy, K = Crack.

3.73
TABLE 3.18. (cont'd)

Base-Line EC Inspections

Type of

Defect(a)

Team

x Y

497 488

319. 316

Number of specimens

Number of Wall-loss

defects

TTS

Detections

False Calls

Nondetections

L

M

H

K

L

M

H

K

L

M

H

K

TSP

Detections

False Calls

N6ndeteCtions L

M

H

K

U-bend

Detections L

M

H

K

False Calls

Nondetections L

M

H

K

(a) L = Light, M = Medium, H Heavy, K = Crack,.

7

35

47

0

0

112

21

4

0

0

0

0(2)

0

0(7)

72(1)

1

0

2

0

0

0(2)

0

0(8)

12

1

0

0

34

45

46

0

0

86

11

4

0

0

0

0(0)

0

0(9)

70,1

1

0

3

0

0'

0(2)

0

0(14)

12

1

0

0

3.74
TABLE 3.18. (cont'd)

Advanced/Alternate Techniques Round Robin

Type of

Defect(a) U

128

UU

Number of specimens 172

Number of wall-loss

-defects,

TeamV

392

245

VV

172

131

W

22

70 131 20

TTS

Detections

False Calls

Nondetect ions

Detections

TSP

False Calls

Nondetections

L

M

H

K

L

M

H

K

L

M

H

K

L

M

H

K

L

M

H

K

L

M

H

K

0

0

0

0

0(3)

8

0

0

0

31

12

8

0

0

7

4

0

0

9

15

26

0

4

72

6

3

0

(b)

(b)

(b)

(b)

(b),

(b)

(b)

(b)

(b)

(b)

(b)

(b)

(b)

25

48

51

0

2

63

8

0

0

0

0

0

0

5(12)

34(l)

0

0(0)

2

0

0

0

0

0

11

1

0

0

11

17

24

0

1

70

4

.5

0

(b)

(b)

(b)

(b)

(b)

(b)

(b)

(b)

(b)

(b)

(b)

(b)

(b)

(b)

(b)

(b)

(b)

(b)

0

9

7

0

0

0

2

2

0

U-bend

Detectio, ns

(c)

(c)

(c)

(c)

(c)

(c)

(c)

(c)

(c)

(c)

(c)

(c)

(c)

(c)

(c)

(c)

(c)

(c)

0

0

0

0

False Calls

Nondetecti~ons

0(3) (b)

0

0

0

0

(b)

(b)

(b)

(b)

(a)

(b)

(c)

L = light, M = medium; H = heavy, K = crack.

Not analyzed because the extent of the inspection

time.

Not inspected.

was uncertain at the

3.75
There were numerous false calls in the support-plate region, only one of

which was at a support-plate intersection. The two base-line teams had more

nondetections than any of the other teams. The majority, of the nondetections

were reported as small defects (<20%).

There were very few wall-loss defects found in the U-bend region. With

the exception of the two grinder marks found by all but one team, the defects

were mostly light fretting wastage at the AVB contact points (one was

classified as medium wall-loss). Since they were the earliest inspections,

Teams X and Y had a tendency to report the small volume indications in the

U-bend as small defects that later proved to be gaps in the copper-rich deposits on the tubes. All of the Team Y indications called were less than

25% wall-loss. Team X's indications, on the other hand, were sized larger; the largest indication was reported as 55% wall-loss. None of the false

positives were in the same specimens.

With the exception of Team V, the conclusions that can be drawn from the

AATRR results are limited, since so few of the specimens were inspected by these teams. Team W did not inspect beyond the TTS. For Teams UU and VV,

the extent of the individual inspections was not given, so the decision was

to report information only for the TTS. Team V's results are similar to those

of the base-line and DAARR teams. However, they reported fourteen 50% throughwall ID defects outside support-plate intersections that turned out to be false

calls.

The probability of detection was determined for both the visual wall-loss

estimates and the metallography data. The visual data provided a larger

number of specimens, while a more precise estimate of wall-loss was obtained

by metallography. Similar POD responses were observed for the Base-Line and

DAARR inspection teams. A plot of the median POD from these seven inspection

teams as function of visual wall-loss is shown in Figure 3.36.

As expected, the median POD for TTS specimens increased with wall-loss and

was 0.94 for specimens in the heavy category, which corresponds to estimated

wall-losses greater than 40%. The POD from the seven teams ranged from 0.88

to 0.98 for this specimen category, while somewhat greater variations were

observed for the light and medium categories. A slight decrease in POD was

observed when specimens from all regions of the generator were included in

the calculations. However, the magnitude of the decrease may be misleading

since the vast majority of defects in the medium and heavy categories were

found at the TTS. It should also be noted that almost all of the data pertain

to pitting or wastage type defects and the PODs may not be appropriate for

other defect types such as SCC or IGA.

The PODs calculated from the metallography data show similar trends to

those from the visual estimates of wall-loss. Figure 3.37 shows the median POD

from the Base-Line and DAARR inspection teams; the metallography data are

grouped into increments of 10% wall-loss. As before, the POD increased with

wall-loss and the average POD for wall-loss more than 40% was greater than

0.9. Below 40% wall-loss, definite team-to-team variations in the POD curves

were observed. However, there are inadequate data in each size category to

3.76
1.0-

0.9- 7<

0.8- 7...

0.7 /

0.6

a/

0.5-

0.4

0

0.3-

0.2 Top Tube Sheet

/....... All Regions

0.1

0.0,

Light Medium Heavy

Visual Wall Loss

FIGURE 3.36. Median Probability of Detection of DAARR and Base-Line

Inspection Teams Based on Visual Determination of Wall-Loss

assess the statistical significance of these variations. An apparent

improvement in the POD curve of one AATRR inspection team was observed, as

shown in Figure 3.38. The POD curve increased more rapidly than the typical

Base-Line or DAARR team and the average POD was higher above 40% wall-loss.

This team utilized an automated data screening system to sort through their

data prior to analysis by a human. This was probably, in part, responsible

for the better POD performance.

Figure 3.39 gives a plot of the individual POD performance for the DAARR

and Base-Line teams for each 10% wall-loss increment. To estimate the lower

bound POD, an approximate 90/90 lower tolerance limit (LTL) was computed (curve

in Figure 3.39) from the teams participating in the DAARR and Base-Line

inspections. These teams are assumed to be typical of the total population

of teams performing ISI; therefore, if each team in this population had

inspected the same set of tubes, we can be about 90% confident that about 90%

of the individual team POD values would be above this LTL. The dashed segment

of the curve indicates that the number of specimens with deep through-wall

degradation was inadequate to provide a meaningful estimate of the LTL. Thus,

the LTL at 65% wall loss is extended as a conservative approximation of the

LTL (i.e., the POD is assumed to either increase or stay the same as wall-loss

increases). This information was developed as an appropriate lower bound

estimate of POD for field EC inspection and was used to validate the selection

POD curves used in the evaluation of various sampling/inspection schemes

described in Section 3.6.

3.77
C

0

0

C0

1.0

0.9

0.8

0.7

0.6

0.5

0.4

0.3

0.2

0.1

0

/

/

-'

7

T.p 'lube Sheet

. - . - . All Regions

10 20 3o 40 5o 6o 7o 80 90 100

% Metallography Wall Loss

Median Probability of Detection of DAARR and Base-Line

Inspection Teams Based on Metallographic Determination of

Wall Loss

FIGURE 3.37.

0.9-

0.8-

6 0.7

.2 0.6-

0 0.5-

0 0.4-

-o

0

CL 0.3,

0i2-

0.1"

ii

~1' I

I

/

1\ll lHgioII//

0 .0 'i 1 I .. I - I 

10 20 30. 40 50 60 70

% Metallography Wall Loss

FIGURE 3.38. Probability of Detection of an AATRR

Based on Metallographic Determination

3.78

80 90 100

Inspection Team

of Wall Loss
1.0

oC08 0 8

0

Q) A

9D0.6-

0 a B

A 03 0 C]

= 0.4 D

umE

0a 0.2-

&Y

* * -90/90 LTL

0.0 -

0 20 40 60 80 100

Metallography Wall Loss, %

FIGURE 3.39. Individual POD Performance for the DAARR and

Base-Line Teams for Each 10% Wall-Loss Increment

Comparing the EC and metallography data shows that EC generally tended to

underestimate the defect depth as measured by metallography, especially for

the more severely-degraded specimens. Figure 3.40 shows the relationship

between the EC and metallographic results for a typical inspection team using

standard bobbin-coil techniques. Points plotted along the horizontal axis

are nondetections and points shown on the vertical axis are false calls.

Team-to-team variations were observed, but the general tendencies shown in

Figure 3.40 were apparent for all of the Base-Line and DAARR inspections.

Wide variations in the EC depth estimates were reported for specimens with

similar wall-loss and also for the same specimen but from different teams. A

better correlation between the EC and metallography wall-loss was observed

for one of the AATRR teams, as shown in Figure 3.41.

Linear models were fit to the team's sizing data using an algorithm

designed for truncated data (Aitkin 1981). The perfect sizing relationship

is described by an intercept of 0.0 and a slope of 1.0. A measure of how well

the model fits the data is obtained from the correlation coefficient, R2 . Summary statistics for the linear sizing model for all teams is given in

Table 3.19. The intercept, slope, sizing error estimate, number of defects

and R2 are listed.

3.79
100

90

80

70 In

0

"J 60

50

0)

E 40

30

20

10

7'

X

x

x

X z0

0 10 20 30 40 50 60 70

Metallography Wall Loss

80 90 100

FIGURE 3.40. Typical Plot of EC Estimated Flaw Depth Versus

Actual Depth from Metallographic Analysis

1001-

90

80

70

60

x

0

E

C',

X

- x

X4< x< x x

,,, . xT ,.,x ý 50' x x x / x 1< x x x

40- x xx X, x

30

20 x/X." x/ 20 •

10-

0 10 20 30 40 50 60 70 80 90 11

Metallography Wall Loss

FIGURE 3.41. Plot of Best Observed EC Estimated Flaw Depth

Versus Actual Depth'from Metallographic Analysis

3.80
Table.3.19 shows little consistency in the estimates of the intercept. .. For the Base-Line and DAARR teams, it varies from 6% to 24%, which represents

a significant sizing bias. The slopes, varying from 0.38 to 0.61, are

reasonably consistent. The sizing errors are also consistent, 14.67% to

16.69%, with the exception of Team Y at 21.11%. These errors are large; a one

standard deviation (error estimate),range about a defect size estimate is at

minimum 30% of wall thickness. Although Team V's fitted intercept and slope

were not significantly different from some of the other teams, their dispersion

was the lowest (except for Team U with only 13 defects >20%) and their R

value was the highest of any team. Team V employed two inspection techniques

that probably explain, in part, their improved sizing performance., First,

special frequency mixes were developed specifically for the Surry generator

and used to enhance the signal-to-noise ratio by suppressing the effects of

dents, support plates, and copper deposits. Second, rotating EC and focused

ultrasonic probes were used to reinspect 22 of the tubes previously inspected

by conventional bobbin-coil EC. The wall-loss determined from the supplemental

inspection was then compared to the bobbin-coil results and a correction factor

computed. In this case the bobbin-coildata was adjusted by +15% to correct

for an apparent underestimate of defect depth. Both of these factors, namely,

special frequency mixes and augmentation of the EC/bobbin-coil data, apparently

were responsible-for the improved sizing accuracy.

TABLE 3.19. Sizing Regression Summary Statistics

Measurement

Team Intercept, % Slope Error, % n R2 Truncation, %

A 12.05 0.49- 16.48 64 0.21 20

B 17.58 0.38 14.67 52 0.13 20

C 24.17 0.40 15.60 62 0.16 20

D 8.96 0.61 16.69 66 0.31 20

E 20.27 0.45 14.81 64 0.24 20

X 9.43 0.49 15.82 58 0.21 0

Y 5.98 0.61 21.11 69 0.26 10

U -4.21 0.84 7.62 19 0.85 20

UU 35.33 0.19 17.80 31 0.04 20

V 14.30 0.64 10.59 73 0.57 20

VV 19.76 0.29 17.94 33 0.09 0

W -33.26 1.40 23.34 14 0.33 10

3.5.3.3 Factors Influencing EC Inspection Reliability

The EC data and reports obtained from the various inspection teams were

examined to determine the cause of the lack of detection for some severely

damaged specimens and the wide variations in sizing. Defect indications that

were missed by one or more of the inspection teams were subsequently found on

the EC data tapes when reanalyzed by a PNL analyst (Bradley et al. 1988).

This indicates that the analysts either bypassed the indication or, for some

reason, did not interpret the generally complex EC signals to be defects.

Interpretation of the complex EC signals also appears to be the primary source

3.81
of the large variations in reported, defect depths for the same specimen.

Specific causes that were identified include: 1) multiple defect signals within the degraded region, 2).different procedures and locations used to

measure the phase angle wit hin dist6rted- signa~s, anJ3); the frequency used

to measure the defect depth. Mostof'the large var-iations in reported defect

depth for a given specimen could'belindividually explained by one or more of

the above causes.

It was also noted that complex EC signal shapes usually coupled with a

low signal-to-noise ratio were general characteristics of specimens with large

variations in the reported defect depths.

The EC data from TTS specimens with metallographic data available were

reanalyzed to determine if a consistent analysis method and or frequency would

improve the correlation between EC and metallography wall-loss. The EC data

tapes from one of the DAARR inspection teams were analyzed at 400 kHz,

100 kHz, and, a 400/100 kHz mix. The defect signals were measured at all three

channels by always taking the location within the signal which produced the

deepest wall-loss. This location usually was a tangent point of a deflection

on the vertical channel of each output. A comparison of the resulting EC

depth estimates with the metallographic data is shown in Figure 3.42. No

definite improvement in the variability of the .EC depth estimates is seen in

the three sets of data, although a consistent analysis method was employed.

The effect of defect type and distribution on the EC sizing variability

was evaluated by dividing the TTS specimens into defect categories consisting

of wastage, pitting and wastage, and isolated pitting. The pitting and wastage

specimens were subdivided into two categories depending on the circumferential

extent of the defects in the transverse sections. Degraded regions (not

necessarily continuous).were distributed around more than one-half of the

tube circumference for Category 1 and less than one-half of the tube

circumference for Category 2. The resulting correlations between the reported

EC wall-loss from the Base-Line and DAARR inspections and the metallographic

data are shown in Figures 3.43 and 3.44. As can be seen, definite improvements in the correlations were not obtained. The simple defect types, wastage and

isolated pitting, appear to provide a slightly improved EC estimate of wallloss. However, the number of specimens in these categories is limited and

the differences may not be significant,

Deposits on the OD surfaces were found to be the primary cause of the

false positive calls made by the inspection teams. :Most of the false calls

were located in the U-bend and TSP regions of the generator, and tubes from

these locations were generally coated with a Cu-rich deposit described above.

Examination of the EC data showed that two types of defect indications were

responsible for the false calls. The first type of indication appeared as a

shallow OD defect and was most easily detected at 100 kHz frequency. After

removal from the generator, correlations were observed between this type of

EC signal and locations where there was a lack of-, or an interruption in the

deposit (Bradley et al. 1988). The, second type of indication was

characteristic of a permeability variation (PV) signal]and was reported as an

ID defect. Removing the deposits. by chemical ,cleaning eliminated both signal

types from most of the specimens, although a small reesidual PV-t-ype signal

-3.82
0

-J

w

0 20 40 60 80 100

Metallography Wall Loss, %

a) 400 kHz

0

0

w

0 20 40 s0 80 100

Metallography Wall Loss, %

b) 100 kHz

100

so

0

-J

s 6o

- 40

O20

0 20 40 60 80 100

Metallography Wall Loss, %

c) 100/400 kHz

FTGURE 3.42. Comparison EC Data Reanalyzed at Various

Frequencies With Metallography

3.83
100

80

(I)

0._j

-60

" 40

2

O 20

w

0

0 20 40 .- 60 80

Metailography Wall Loss, %

100

a) Wastage Defects

80

u)

CO

0

-j

CO

0i

100

80

60

40

20

0

0 20 40 60 80

Metallography Wall Loss, %

100

b) Pitting Defects

FIGURE 3.43. Correlation of EC Estimated-Wall-Loss (Base-Line

and DAARR Teams) With Metallography

3.84
10I-

~0

C

0

wU

100

80

60

40

20

0

0 20 40 60 80 100

Metallography Wall Loss, %

a) Pitting/Wastage Less Than Half Circumference

100

-0

0-

(D,

0

VU

80

60

40

20

0

0 20 40 60 80 100

Metallography Wall Loss, %

b) Pitting/Wastage More Than, Half Circumference

FIGURE 3.44. Correlation of EC Estimated Wall-Loss (Base-Line

and DAARR Teams) With Metallography

3.85
remained in a few specimens. Visual inspection of all specimens and

metallography on selected specimens found no evidence of defects. Superficial

fabrication marks and spots without the thin oxide coating on the tube OD

were noted in the inspection data because of possible correlations with

interruptions in the Cu-rich deposits.

A large number of the EC signals observed on the inspection data tapes

were due to interruptions in the Cu-rich surface deposits. These signals

tended to be classified as defect indications by inspection teams depending

upontheir amplitude, shape and location within the generator. Figure 3.45

provides a typical example of an EC signal caused by the Cu deposits. The

signal shown in Figure 3.45 was interpreted by an inspection team as a defect.

After removal from the steam generator,'this specimen was EC inspected and

then chemically cleaned to remove the Cu-rich deposits. Following chemical

cleaning, another EC inspection was performed. In all cases, the EC signals that could have been interpreted as a defect were no longer visible.

Figure 3.46 shows the post-cleaning EC inspection results for the specimen in

Figure 3.45; note the defect indication has disappeared. Destructive

examination of some specimens with these types of signals confirmed the absence

of defects.

3.5.4 Burst Testing of Service-Degraded Tubes

The objective of the burst-test work was to validate empirical models of

remaining tube integrity developed during Phase I of this program (Alzheimer

et al. 1979). Burst tests were performed on tubes removed from the steam

generator with pitting/wastage type defects. The results of these tests are

described below.

Twenty specimens removed from the hot-leg TTS region of the steam

generator were burst tested to measure remaining tube integrity. Seventeen

specimens with visually-severe pitting/wastage type defects from above the

TTS were tested as well as three specimens taken from the TS crevice with no

observable defects. Undefected portions of tubing were tested so that the

normalized burst pressure of defected specimens could be computed. This was

done to minimize the influence of material property and tube dimensional

variations on burst strength. Thus, the dependence of tube integrity on just

defect severity could be determined.

Prior to testing, specimens were cleaned to remove residual radioactive

contamination, photographed, and dimensionally characterized. A wall thickness

gage was used to estimate the amount of remaining wall thickness for the most

severely-degraded regions of each tube. The wall-thickness measurements were

recorded on the specimen photographs.

After burst testing, the length and depth of defect that caused each

specimen to fail was estimated by comparing the failed tube with the pretest

photographs that show the remaining wall-thickness measurements. In most

instances, the depth of defect causing tube failure corresponds to the region

of maximum pitting/wastage. The burst-test results (see Table 3.20) show

that even deep pitting/wastage did not significantly reduce tube strength.

All tubes leaked or burst at levels several times the highest pressure

3.86
FIGURE 3.45 U-Bend Specimen 568 (R4 C3) Showing a Defect-Called Signal

in U-Bend Area Which Turned Out to be an Interruption in the

Copper-Rich Surface Deposits

3.87
I

la bti.

H I~ IVERT *[.C,4 2 VERT CHOANN~EL W4O-

- ---------

C1- 140 I DIFF

FREQ - 400 kHz

SPAN - 272

ROTATION- 358 DEG

CHANNEL 2 H

FREQ - 180 kHz

SPAN - 494

ROTATION - 7 DEG

C1-HEL 3 DIFF

FREQ - 240 kHz

SWAN - 57-2

ROTATION- 8 DEG

CHANNEL 4 DIFF

FREQ - 5W8 kHz

SPAN - 314

ROTATION - 2 DEG

,MIX 1 (5) <-

CHANNELS 12

SPAN - 1824

ROTATION- 0 DEG

MIX 2 (6) <-

CHANNES 13

SPAN - 1824

ROTATION - 0 DEG

~IIIO ORow~ 4c,t 3O

CH._0 IVERT CH 2 VERT j CHA1NNEL NO - 3 jIo a Row 4 cot. '34 !

1 - i -

CHANNEL 1 DIFF

FREQ - 480 kHz

SPAN - 272

ROTATION- 358 DEG

CHANNEL 2 DIFF

FREQ - 180 kHz

SPAN - 494

ROTATION- 7 DEG

CHANNEL 3 WM

FREQ - 240 kHz

SPAN - "7

ROTATION - D0EG

CHANEL 4 DIFF

FREQ - 500 kHz

SPAN - 314

ROTATION- 2 EG

MIX 1 (5) <-

CHANNELS 12

SPAN 1024

ROTATION- 0 OEG

MIX 2 (6) <-

0CHANELS 13

SPAN - 1024

ROTATION- 8 DEG

m

i -- " _

1% ~

Th

FIGURE 3.46. Specimen 568 (R4 C3) After Cleaning

3.88
TABLE 3.20. Surry Tube Burst Data

'00

Specimen

No.

657B

794B

797B

601C

603C

635C

657C

794C

797C

615C

628C

642C

661C

712C

790C

792C

795C

799C

812C

826C

Test

Temp. ,-F

550

550

550

550

550

600

550

600

550

600

6o0

600

600

600

600

600

600

600

600

600

49

55

51

69

78

50

37

65

36

75

38

57

38

29

51

44

57

Max. P/W

Depth, %

Max. Flaw.a) Depth, %

24

55

43

69.

78

50

24

52

36

75

38

57

38

29

8 0 (b)

22

43

Max. Flaw(a)

Length, in.

0.06

0.15

0.10

0.15

0.17

0.10

0.06

0.10

0.04

0.11

0.06

0.05

0.07

0.04

0.53 (b)

0.08

0.10

Normalized

Burst Burst Pressure,

Pressure, psig _ AP/APo_

11,165 1.00

10,720 1.00

11,765 1.00

10,540 0.94

9,790 0.87

10,825 0.97

8,835 0.79

7,175 0.67

11,110 0.94

10,152 0.91

10,078 0.90

10,498 0.94

9,195 0.82

10,652 0.95

10,300 0.92

10,128 0.90

10,800 0.96

6,930 0.62

9,988 0.89

10,485 0.94

(a) Corresponds to

(b) Max. depth and

max. depth and length of

length of an axial crack.

P/W at failure location.
attainable in a main-steam-line-break accident (about 2600 psi). This was

because of the short length of these defects both axially and

circumferentially.

Figure 3.47 shows a plot of the calculated normalized burst pressure

against measured values. Excellent agreement was obtained between calculated

and measured values for the Phase I EDM notch empirical model. Similar results

were obtained for the uniform-thinning model, but the elliptical-wastage model

predicted burst pressures substantially below actual values. This model tended

to overpredict burst pressure because it was developed from burst-test results

on tubes with long axial elliptical wastage. All data points, except one,

were within 410% of perfect agreement. The one data point outside the *10%

range was from a tube in which failure was caused by an OD initiated SCC.

The actual burst pressure of this tube was about 20% greater than the predicted

value. This result was partially due to the method employed for characterizing

the crack dimensions. The cracked portion of the tube was, in fact, composed

of several small closely spaced cracks. To conservatively characterize the

flawed area, an overall crack length consisting of the sum of the smaller

cracks was used. The crack depth was estimated from post-test fracture surface

measurements and a metallographic section taken through the center point where

the crack had penetrated the tube wall. It is clear that the EDM notch

empirical equation gave conservative, results for the axial SCC when bounding

dimensions were used to characterize its size.

The burst-test results illustrate how bobbin-coil EC depth measurements

provided a conservative estimate of remaining tube integrity even with

significant undersizing. Figure 3.48 gives a plot of burst pressure versus

median EC estimated defect depth for the Base-Line and DAARR teams. The boxed

area in the lower left-hand corner of Figure 3.48 defines a zone bounded by the

present 40% plugging limit and three (the present factor of safety specified

in Regulatory Guide 1.121) times normal operating pressure differential.

Note that none of the data fall within the nonconservative zone. This was

because of the short axial extent of the pitting/wastage. Large uncertainty

in the EC depth estimate can be tolerated when the true defect length is short

(<0.25 in.). Thus, the 40% plugging limit is conservative for these defect

types. Accurate measurement of defect depth becomes much more important when

the defect length is greater than about one tube diameter.

3.6 INSERVICE INSPECTION SAMPLING PLAN EVALUATION

A major objective of the statistical work was to evaluate and compare

candidate sampling/inspection schemes for ISI of steam generator tubes. One

criterion for comparing'sampling/inspection schemes is the probability of

detecting and plugging a tube that has a flaw with a specified through-wall

depth. For a single tube, this probability is a function of two other

probabilities: (1) the probability of detection, which is the probability of

observing a positive eddy-current reading, and (2) the probability of plugging

(POP), which is the conditional probability that a positive EC reading will

exceed the plugging limit and result in plugging or repairing the tube. Both

the POD and POP are functions of the true size and type of flaw. They also

depend upon the capability and reliability of the inspectors and their

equipment.

3.90
V)

CO

(L

4-

CO

N

0

Z

a.)

FIGURE 3.47.

0.0 .0.2 0.4 0

Measured Normalized B

Calculated Normalized Burst Pr

vs Measured Normalized Burst F

Removed From the HLTTS Region

.6 - 0.8 1.0.

urst Pressure

ressure From EDM Slot Equation

Pressure for Tubes

With P/W Defects

aU)

L.

C')

CO

U)

I... 0~

4-U)

I...

12000

10000-

8000-

6000-

4000

2000-

0

0 &0

0 d MO0 0

0

0

0

1.

4. I

0

I -I I I

20 40 60 80

Median EC Indicated Depth, %

100

FIGURE 3.48. Burst Pressure vs Median Eddy-Current Depth for

Tube Segments From the Surry Steam Generator

3.91
The Surry generator inspection data were used to develop statistical

models of POD and POP as functions of true flaw size for flaws due to wastage

and pitting. Because multiple inspection teams were involved, the statistical

modeling yielded a range of estimated POD and POP values for each specified

flaw size. These'ranges of values were utilized with probability theory and

Monte Carlo simulation techniques to evaluate and compare sampling/inspection

schemes. This section provides a brief overview of the statistical approach,

results, and conclusions. A comprehensive report has been prepared (Bowen

et al. 1989) that provides a detailed account of this work.

3.6.1 EC Sizing Error and Probability of Plugging

For the purpose of characterizing EC sizing error and estimating POP

values, results from the destructive metallographic analyses were matched

with the EC inspection results for Surry generator tube segments. EC

inspection data are present for 12 teams. Because POP is conditional upon a

positive EC indication, the data set for each team was reduced by removing

data pairs with no positive EC indication.

The statistical methodology used to model EC sizing error is based on the

following definitions and assumptions. For a particular team and a particular

"flaw," define

X = "True" Maximum Flaw Depth Determined from

Destructive Metallographic'Analysis (DMA)

Y = Observed Flaw Depth as Determined !from

EC Inspection

Assume that Y can be modeloed as'a-simple linear function of X but with random

measurement error e, so that

Y = a + bX + e (3.1)

That is, at a specified value of X, the distribution, of Y is assumed to

have true mean a + bX and variance Var(e), where a, b, and Var(e) are unknown

parameters that must be estimated for each team from the inspection data.

The iterative algorithm presented by Aitkin (1981) was used-to fit the

model in'Equation (1) to the inspection data for each team. This technique

for fitting models to data with some censored observationstproduces estimates

of a, b, and Var(e) with desirable statistical properties. Also, a more

sophisticated model was fitted to the combined data from the five DAARR teams

that incorporates team-to-team variability into the estimate, of "Var(e)."

3.6.2 Probability of Exceeding EC Plugging Limit (PEL)

It is of interest to estimate the probability that a non-zero EC value

will exceed a specified "plugging limit" T for a flaw with "true" depth X.

It is also of interest to'determine'an EC plugging limit such that the

probability of plugging or repairing defective tubes is acceptably high. The

3.92
fitted linear models described previously provide a means for achieving, these

objectives.

For a particular fitted model and a specified EC plugging limit T, the

probability of exceeding the EC plugging limit for atube with a positive EC

indication and a flaw with "true" DMA depth X can be evaluated as follows.

The predicted mean EC value is computed from the'formula Y = a +bX with

estimates of a and b substituted. The variance of the distribution of EC

values at X is the sum of the estimate of Var(e) and the variance of the

predicted EC mean value. The probability that an' observed'EC value will

exceed T is estimated from the normal distribution with mean and variance set

equal to their estimated values.

In this study, a defective tube was defined as one having a flaw with true

through-wall depth of at least 75%. If it is assumed that the EC plugging

limit is 40%, the estimated PEL values (probability that Y >40% given that Y

>0%) for six of the fitted-models are displayed in Table 3.21. For example,

when a tube has a flaw with true depth X = 75%, and a non-zero EC value has

been observed by a team like one of theDAARR teams, the probability is 0.73

that the observed EC value will be greater than T.= 40%. Clearly, a range of

sizing capabilities is represented by the six models.

,TABLE 3.21. Estimates of PEL for X - 75% and T = 40%,

Round Robin 0O.733 Team UU 0.695

Team Y 0.709 Team X 0.644

Team V `0.986 Team VV 0.471

It is also of interest to determine for each model the value of the EC

plugging limit:T required to:achieve a specified PEL when X = 75%. For

example, suppose that it is desirable'to have PEL = 0.9,when X =375%. The

required T values, estimated from the normal distribution, are displayed in

Table 3.22 for the six models. If, for example,, the DAARR Teams or teams UU,

X, or Y are representative of the teams that currently perform inservice

inspections, then the EC'plugging limit would have to be lowered to about

25%.

TABLE 3.22. Plugging Limit T Required for PEL = 0.9 When X = 75%

Round Robin' 29.2 Team UU 25.2

Team Y 26.6 Team X 25.5

Team V 49.8 Team VV 14.2

If Team VV is the "norm," then a plugging limit at about 14% would be

required. However, if all teams could perform like Team V, the plugging limit

could be increased to 50%.

3.93
The sizing capabilities of the teams can be compared by comparing the

fitted models or by comparing PEL values. For example, Table 3.23 displays PEL

values for three of the models with plugging limit values T= 30%, 40%, and

50%.

TABLE 3.23. Estimated PEL Values for Three Plugging Limits

Plugging True Round Robin

Limit, % Depth, % Teams Team UU Team V

T =30 0 0.23 0.65 0.05-

20 0.41 0.72 0.36-

30 0.51 0.75 0.61'

50 0.72 0.80, 0.94,ý

75 A0.89 .0.85- 1.00

T = 40 0 0.09 0.47 0.01

20 0.20 0.53 0.09.

30 0.28 0.56 -0.,25

50 0.48 0.63 0.73

75 0.73 0.70 0.99

T =50 0 0.03 0.29 0.00

20 0.08 0.33 0.01

30 ,0.12 - 0.36 0.05

50 0.26 0.42 0.36

75 0.51 0.50 0.90

Note that Team UU always has higher PEL values for nondefective tubes

(X <75%) than the other teams. If the plugging limit is set at T = 30% so

that the DAARR teams and Team UU have a high PEL when X = 75%, all teams have

PEL >0.5 when X >30%; and Team UU would tend to plug most of the tubes with

positive EC indications. If T is increased to 50%, Team V has a high PEL

when X = 75%, and would not be likely to plug tubes with true flaw depth

X <50%; but the DAARR teams and Team UU have low PEL values when X = 75%.

3.6.3 Probability of Detection (POD)

Estimates of POD were computed by matching EC inspection results with the

results from both the visual-inspections and destructive metallographic

analyses. For each "true flaw size" category, the number of non-zero EC

indications divided by the'total number of flaws was used as a POD estimate.

A POD curve (i.e., a plot of estimated POD vs "true size") was constructed

for each inspection team, and an overall POD curve was constructed by combining

the data from all teams. Some of these curves were presented in Section

3.5.3.2, and the ranges of estimated POD values were used as a basis for

evaluating and comparing the performance of sampling/insp3ction schemes.

3.94
3.6.4 Evaluation and Comparison of Sampling/Inspection Schemes

There are two basic strategies for selecting tubes from a generator for

inservice inspection. Either all tubes are inspected (100% inspection) or a

sample of the tubes is selected for inspection. Although there are many

possible sampling schemes that could be applied, one particular type of

sequential sampling/inspection was identified as most appropriate and was

evaluated and compared with 100% inspection.

With 100% inspection, all defective tubes in a generator will be

inspected, and the joint probability, p, of detecting and plugging an

individual defective tube is the product of the POD and the PEL for a defective

tube. That is,

p = POD(PEL) = Pr(Detect and Plug I Defective) (3.2)

To be conservative in this study, POD = 0.9 was assumed for flaws large

enough to classify a tube as defective. Thus, p can be computed from

Equation (2) for a specified PEL value. When there are n defective tubes in

the generator, the probability that k or more defective tubes are plugged after

inspection is computed from the cumulative binomial distribution with

parameters n and p.

When a sampling scheme is applied to select tubes for inspection,, there

is no guarantee that all defective tubes in a generator will be inspected.

Thus, the probability of detecting and plugging a defective tube is a function

of the probability that the defective tube will be inspected. Without further

assumptions about the distribution of defective tubes in a generator or

supplementary sampling/inspection, Equation (2) would be multiplied by the

probability• of inspection. For example, if POD = 0.9, PEL = 0.7, and if 3%

of the tubes are randomly selected for inspection, then the probability of

inspecting, detecting, and plugging an individual defective tube is

p = 0.9(0.7)(0.03) = 0.0189. With 50% random sampling, p = 0.9(0.7)(0.5) =

0.315. These values compare with p = 0.9(0.7) = 0.63 for 100% inspection.

By making assumptions about the distribution of defective tubes and by

considering a particular type of sequential sampling/inspection scheme, the

effectiveness of sampling/inspection relative to 100% inspection can be

improved considerably over the completely random sampling implied above.

In the analytical portion of this study, it was assumed that defective

tubes tend to occur in "clusters," which are groups of defective and degraded

tubes. For the purpose of evaluating and comparing sampling/inspection

schemes, a "minimum" cluster was assumed, which is a defective tube surrounded

by degraded but not defective tubes in the following pattern:

0

OXO

0

where X denotes a defective tube and 0 denotes a degraded but not defective

tube.

3.95
It is recognized that in a real generator, clusters could be shaped

differently than the one shown above, and could be different sizes, and could

include more than one defective tube. The above cluster configuration was

selected for evaluation purposes because it would be harder to detect than a

larger cluster or a cluster with more than one defective tube. It is also

recognized that in some cases defective tubes may be isolated. As a result,

in the simulation analysis portion of the study (see Section 3.6.5), various

distributions of degraded and defective tubes were examined to evaluate other

conditions of clustering ranging from nearly isolated defectives up to a single

large cluster.

The sequential sampling/inspection scheme that was chosen for evaluation

is assumed to proceed as follows:

a) The initial sample is selected according to a systematic sampling plan

that consists of a specified percentage of the tubes in a generator, and

each tube in the sample is inspected.

b) When a positive EC indication is observed, inspection continues in the

region immediately surrounding the suspect tube until a 2-tube wide

"buffer zone" is observed, which is composed of tubes with no positive

EC indication, and which completely surrounds the tube(s) with positive

EC indication(s).

c) In steps a and b, each tube with an EC indication that exceeds the

plugging limit will be plugged or repaired.

By assuming the above cluster configuration, it is possible to define a

20% systematic sampling plan for step (a) that would include exactly one tube

from each cluster. It is also possible to define a 40% systematic sampling

plan that would include exactly two tubes from each cluster. Then, if the

defective tube in a cluster is not included in the initial sample, there is a

chance that the degraded tube(s) will produce a positive EC indication that

will trigger additional inspection (step b), which will include the defective

tube.

Thus, the probability of inspecting and detecting the defective tube,

denoted by PI&D, is a function of the POD for degraded but not defective

tubes, denoted by POD(deg), as well as the POD for defective tubes (which is

assumed to be 0.9). Specifically, with 20% systematic sampling at the first

stage,

PI&D = 0.18 + 0.72 POD(deg) (3.3)

and with 40% systematic sampling at the first stage,

PI&D = 0.36 + 1.116 POD(deg) - 0.54 [POD(deg)] 2 (3.4)

The joint probability of detecting and plugging a defective tube is given by

p = PI&D(PEL) (3.5)

3.96
If there are n-defective tubes (n clusters as defined above) in a

generator, and if the clusters are assumed to be far enough apart so that the

inspection results from one cluster do not affect those of other clusters,

the probability that k or more defective tubes are plugged after inspection

is computed from the cumulative binomial distribution with parameters n and p.

For evaluating and comparing the sequential sampling/inspection schemes

and 100% inspection, PEL values ranging from 0.50 to 1.0 were considered (the

PEL estimates from the six fitted models range from 0.471 to 0.986), together

with values of POD(deg) ranging from 0.5 to O.7'. For the sequential

sampling/inspection schemes with 20% or 40% initial sampling, values of p

were computed from Equation (3.5). The resulting values of p are displayed in

Table 3.24. For 100% inspection, values of p were computed from Equation (3.2)

by assuming POD = 0.9.

TABLE 3.24. Values of p Computed for a Range PEL and POD(deg) Values

Assuming That POD = 0.9 for Defectives

PEL

0.50 0.60 0.70 0.80 0.85 0.90 0.95 1.00

20% Sequential

0.50 0.2700 0.3240 0.3780 0.4320 0.4590 0.4860 0.5130 0.5400

POD(deg) 0.60 0.3060 0.3672 0.4284 0.4896 0.5202 0.5508 0.5814 0.6120

0.70 0.3420 0.4104 0.4788 0.5472 0.5814 0.6156 0.6498 0.6840

40% Sequential

0.50 0.3915 0.4698 0.5481 0.6264 0.6656 0.7047 0.7439 0.7830

POD(deg) 0.60 0.4176 0.5011 0.5846 0.6682 0.7099 0.7517 0.7934 0.8352

0.70 0.4383 0.5260 0.6136 0.7013 0.7451 0.7889 0.8328 0.8766

100% Inspection

0.45 0.54 0.63 0.72 0.765 0.81 0.855 0.90

Note in Table 3.24 that when POD(deg) = 0.7, the sequential sampling/

inspection scheme with 40% systematic sampling at the initial stage yields

values of p that are close to those obtained for 100% inspection. In fact,

by setting Equations (3.3) and (3.4) equal to 0.9 and then solving for

POD(deg), a value of POD(deg) = 1.0 would be required for the 20% sequential

plan to perform exactly like 100% inspection, whereas POD(deg) = 0.77 would be

required for the 40% sequential plan to perform exactly like 100% inspection.

Numerous tables were computed for each of the three plans which display the

probability of leaving a specified number of defective tubes unplugged after

inspection when there are n (n ranged from 2 to 20) defective tubes in the

generator prior to inspection. These results indicate that when POD(deg) is

approximately 0.7, the 40% sequential plan nearly duplicates the performance

3.97
of 100% inspection (Bowen et al. 1989). It must be emphasized, however, that

these results are dependent upon the cluster assumption discussed previously.

3.6.5 Monte Carlo Simulation Analysis

To supplement the analytical results, it was desirable to evaluate the

performance of the sampling/inspection schemes when the assumption of isolated,

perfectly-shaped-cluster sets does not hold. This was accomplished by

application of Monte Carlo simulation techniques. A computer program was

developed that simulates 100% inspection and the sequential sampling/inspection

scheme with 20% or 40% systematic sampling at the initial stage.

A generator with the same number of tubes as the Surry generator (i.e.,

3,388) was assumed, and five different tube maps (shown in Figures 3.49 to

3.53) were considered. As shown in the analytical evaluation, the

effectiveness of a sampling/inspection scheme depends on probability of

detection (POD), sizing capability, and plugging limit. Thus, to simulate

the inspection of a tube, it was necessary to input: 1) a POD model that

expresses POD as a function of true flaw size X, 2) an EC sizing model that

expresses the expected (or mean) EC size as a function of true flaw size X

and also provides the standard deviation of individual EC values about the

mean value, and 3) an EC plugging limit such that a tube with an EC reading

that exceeds the plugging limit will be "plugged." It was of interest to

study how changes in any or all of these factors would affect the performance

of each sampling/inspection scheme. To accomplish this, several different

POD curves were considered. Figures 3.54 and 3.55 show the most important

POD curves utilized.

Each POD curve defines a POD value for'any true flaw size from X = 0 to

X = 100% through-wall. The POD Curves 1 and 2 in Figure 3.54 were chosen to

represent lower (Curve 1) and upper (Curve 2) bounds on the POD estimates

obtained from the Surry inspection data prior to the final POD analysis

presented in the Task 13 Report (Bradley et al. 1988). Note, however, that

Curves 1 and 2 do not include a false call probability; that is, in Figure

3.54 both curves have POD = 0 when the true flaw depth is X = 0. Although

this zero false call probability is not realistic, false calls can only improve

the effectiveness of the sequential sampling/inspection scheme. Thus, an

evaluation of effectiveness with a zero false call probability will tend to

be conservative.

The POD Curves 4 and 5 in Figure 3.55 are refinement of Curves 1 and 2

based on later (but not the final) POD estimates. The differences are as

follows. Curve 4 rises from POD = 0.10 to POD = 0.90 on the interval X = 20

to X = 60; whereas Curve 1 does not reach POD 0.90 until X = 75. On the

interval X = 0 to X = 40, Curves 2 and 5 are identical. For X > 40, however,

Curve 5 reaches a maximum POD = 0.95 at X = 75 and then remains at POD = 0.95

for X > 75; whereas Curve 2 increases faster than Curve 5 and reaches a maximum

POD = 1.0. Thus, Curve 5 is a less optimistic lower bound on POD than Curve

I. Note also that Curves 4 and 5 have a zero false call probability; that

is, POD = 0 at X = 0.

3.98
Percent Flaw Depth

o < 20

* 20 - 49

50 - 74

75 - 100 ..... ... ..

0

.... I

-NOZZLE FIGURE 3.49. Tube Map 1 MANWAY-
Percent Flaw Depth

C) < 20

0 20 - 49

..............

50 - 74

75 - 100

w

i m I

0

* m

-NOZZLE FIGURE 3.50. Tube Map 3 MANWAY-
Percent Flaw Depth

o < 20

.20- 49

50 - 74

75 - 100

•..com .....

cam ew ......

°D

emx

CDC

S

CM

C

g~.a

S..

S.

S..

00w

0e00c

e~wm 

coo0

S

(D)

Ce

COs

.. . . . . . . . .. . • on0m

.eD~iYX!. C .. 19C ...........

so

CM 0(!M CD(. 00 ..... ..... 00C00D......

-NOZZLE FIGURE 3.51. Tube Map 6 MANWAY-
Percent Flaw Depth

CD < 20

@20 - 49

50 - 74

@75 - 100

0

0

0

4*1

a•

0

emr

9000000oUD

em ocs2U Q

(Dn WX 6~oo

0w 

o

(

0 (

eve. .................... 0w.........

o

.0(:.

00

-NOZZLE FIGURE 3.52. Tube Map 8 MANWAY-
percent Flaw Depth

(D <20

e20- 49

S5 - 74

* 75 - 100

. . . . . . . . . .

0

O so08

0

0

0

0

('a

0

C- o

*4o

0 00 • (

0D oD •

em WDZ8 00(

a sc* 00

• swa • ftovm. 6 * oo eogo 0 0OO SOO O

0 OO 0 0 a w OseX•~ 86.. .O00• Cr 9 Di *(r0 S00 s 9 Do04

0 (D coos 0 0 (. AS

0 0 o m e * • G eomoooe r ooosm

.0 000 *@S* C 0 0 *C40

0O* e go * 00 O 000 o

* .eo5ooo•" *S oa5_ to°°o o ceo •.

S

FIGURE 3.53. Tube Map 13 Predicted Surry "True State" MANWAY- -NOZZLE
1.0

0.8

Q)

¢' 0.6

0

0.4

=0

n- 0.2

0.0

0 20 40 60

True Flaw Depth, %

80 100

FIGURE 3.54. Simulation POD Curves 1 and 2

1.0

0

Q)

a)

~0

.0.

0~

0.8

0.6

0.4

0.2

0.i

0 20 40 60 80

True Flaw Depth, %

100

FIGURE 3.55. Simulation POD Curves 4 and 5

3.104
To study the effect of sizing capability on the effectiveness of the

sampling/inspection schemes, two EC sizing models were considered. Model 1

in Table 3.25 is intended to represent the "average" sizing capability of

U.S. inspection teams. Specifically, the parameters a and b were estimated

by averaging the values of a and b (see Equation 3.1) for Teams A, B, C, D,

and X. The standard deviation, SD, was estimated by first averaging the Var(e)

values for these teams and then taking the square root of the average. Model

2 in Table 3.25 is the fitted model for Team V and is intended to represent

an achievable level of sizing performance.

TABLE 3.25. EC Sizing Models Used in Simulations

Model Equation SD* Description

1 14.5 + 0.46(X) 16 Average US Team

2 12.6 + 0.68(X) 10 Team V

*SD .= Standard Deviation

The simulation analysis strategy was to consider various combinations of

tube map, POD curve, EC sizing model, and plugging limit. For a given

combination of these parameters, 25 independent applications of all three

sampling/inspection s 'chemes were simulated. For each combination, results

from the 25 simulated inspections were summarized in tables and plots.

The principal measure of sampling plan performance for detecting and

plugging defective tubes was the sampling effectiveness. The sampling plan

effectiveness was defined as the ratio of the average number of defective

tubes plugged to the total number of defective tubes in the tube map. The

effectiveness parameter provided a means for comparing the plugging capability

of various sampling plans across the different tube maps.

The 'results of the Monte Carlo simulation analysis (see Figures 3.56 to

3.5g) support the conclusions reached from the analytical evaluation and

provide some valuable additional insights. When the isolated clustering

assumption discussed in Section 3.6.4 is approximately valid, the

sequential/40% scheme tends to be nearly as effective as 100% inspection, but

requires substantially less inspection. Note that for conditions of small to

almost no clustering (tube maps 1 and 6), 40% sampling plan is significantly

better than 20% sampling. However, when the defective tubes are in one large

cluster with degraded tubes, all three sampling/inspection schemes are equally

effective for detecting and plugging defective tubes (Figures 3.56 and 3.57).

For the single large cl~uster cases, the results imply that a sequential scheme

with less than 20% initial sampling might be adequate; however, no simulations

were performed to determine how small the initial sample could be before the

effectiven'iss is significantly decreased.

3.105
0.8

U)

uLJ

0.6

0.4

0.;

0 20 40 60 80

Initial Systematic Sample,%

100

FIGURE 3.56. Sampling Plan Effectiveness Versus Systematic

Sample Size for Various Tube Maps -

POD Curve 1/Sizing Model 1

1.0

C-,

a)

0.8

0.6

0.4

0 20 40 60 80

Initial Systematic Sample,%

100

FIGURE 3.57. Sampling Plan Effectiveness Versus Systematic

Sample Size for Various Tube Maps - POD Curve 2/Sizing Model 2

3.106
U)

aLii

0 20 40 60 80

Initial Systematic Sample,%

100

FIGURE 3.58. Sampling Plan Effectiveness Versus Systematic

Sample Size for Tube Maps 1 and 6 -

POD Curve 4/Sizing Model 1

1.0

(n

a)

a) C:

LLJ

0.8

0.6

0.4

0 20 40 60 80 100

Initial Systematic Sample,%

FIGURE 3.59. Sampling Plan Effectiveness Versus Systematic

Sample Size for Tube Maps 1 and 6 - POD Curve 5/Sizing Model 2

3.107
Improving the defect sizing capability to the best observed in the Surry

inspection data yields improved effectiveness of all three sampling/inspection

schemes. (Compare Figure 3.57 to 3.56 and Figure 3.59 to 3.58). Also, with

the improved defect sizing capability, increasing the EC plugging limit from

40 to 50 has little impact on the effectiveness of each sampling/inspection

scheme for plugging defective tubes, but it minimizes the probability of

plugging nondefective tubes. Whereas if the "average U.S. team" sizing

capability observed in the Surry inspection data is assumed, increasing the

plugging limit from 40 to 50 significantly decreases the effectiveness of all

sampling/inspection schemes. (Results not presented here; see Bowen et al.

1989.)

Improving the POD enhances the effectiveness of all sampling/inspection

schemes, and the best performance is achieved when the best POD curve and

sizing model are assumed. However, for the cases and the POD curves

considered, flaws with true sizes less than 20% through-wall do not

significantly contribute to the effectiveness of the sequential schemes (see

Bowen et al. 1989).

The final issue studied in the simulation analysis is the impact of a 0.05

false call probability on the total number of tubes inspected with either

sequential scheme, and whether the additional inspections triggered by false

calls enhance the effectiveness of the sequential schemes. For the sequential

schemes, the resulting false calls triggered inspection of an additional 19%

to 26% of the 3388&tubes in the generator. However, a 0.05 false call

probability does not cause either the 20% or 40% sequential scheme to increase

to 100% inspection. Also, for the cases considered, there is no firm

indication that the false calls improve the effectiveness of the sequential

schemes (see Bowen et al. 1989).

3.7 INTERGRANULAR STRESS CORROSION CRACK ROUND ROBIN

To supplement the NDE validation data obtained from the SGGP, an

additional round robin was performed to provide information on the reliability

of EC techniques to detect and size SCC under-simulated service conditions.

Fifteen tubes with laboratory produced SCC, along with one tube containing a

through-wall SCC that had been previously leak-rate tested, one blank tube

and an ASME flat-bottom hole standard, were assembled into a bundle and sent

sequentially to the J. A. Jones Applied Research Center, Universal Testing

Laboratories/(KWU), Zetec, and Combustion Engineering. It is important to note

that three of these firms routinely conduct inservice inspections of steam

generators. Thus, the results of this round robin should provide an estimate

of field inspections to detect and size SCC. Some of the tubes were coated

with a i- to 2-mil nonuniform layer of copper to simulate the type of deposits

observed on tubes removed from the Surry generator. In other cases the cracked

portion of the tube was covered either partially or completely by an 0.75-in.

thick carbon steel support plate. In these instances the tube-to-tube

support-plate crevice was not filled with magnetite or other typical crevice

deposits, nor was the tube dented in any way. Each participant in the round

robin was required to perform a standard bobbin-coil inspection with 100 kHz

and 400 kHz frequencies and any other frequencies of their choice. Each was

also permitted to inspect the tube bundle with any alternative method desired.

3.108
Typically, segmented bobbin-coil and rotating pancake-coil probes were used

for the alternate inspections.

Figure 3.60 summarizes the POD results for the fourteen teams/techniques

that inspected the tube bundle blind. Due to the limited amount of data, no

attempt was made to determine POD as a function of crack dimensions. However,

it should be noted that in 12 of the 16 tubes the crack depth was greater

than 40% through-wall. To summarize, the average probability of detection of

SCC for teams using bobbin-coil techniques was 0.5. Surprisingly, the

alternate bobbin-coil techniques performed significantly below techniques

using the conventional bobbin-coil probe. The best detection was obtained

from teams utilizing rotating pancake-coil (RPC) or array-coil techniques

either alone or as a supplemental technique for conventional bobbin-coil

inspection. The average POD for teams using rotating pancake-coil techniques

was 0.63.

1.0

' 0.8

0

(D0.6

Co

0,4

..M

0

L)

C) 0.2

0.0

MB MC MD ME MF MH MI MJ MK ML MM MN MO MP

Inspection Team

FIGURE 3.60. Summary of POD Results for IGSCC Round Robin Teams

Teams were asked to characterize detected SCC and provide estimates of the

crack depth and length. Although the amount of data collected during the

mini-round robin was limited, typical results for EC estimated crack depth

and length versus destructive measurements are shown in Figures 3.61 to 3.68.

It is evident from these graphs that the reliability of EC for crack depth

and length was not very accurate or precise. Some teams tended to undersize,

while other teams oversized. Another observation from this exercise was that

the alternate inspection techniques were not significantly better at sizing

SCC then the conventional NDE methods. As a final note, the reader should be

3.109
100

60

80

2 0

i-

-

0

0

FIGURE 3.61.

100.

80

S60-

40-

'0

20°

0

0 20 40 60 80 100

Actual Crack Depth, %

EC Indicated vs. Actual SCC Depth for Team MD - Bobbin

0

0 20 40 60

Actual Crack Depth, %

80 100

FIGUrE 3.62. EC Indicated vs. Actual SCC Depth for Team MF - Bobbin

3.110
100

- 8080

-• 6060

04

20

- 20 0

0

0 20 40 60

Actual Crack Depth, %

80 100

FIGURE 3.63.,

:.100

EC Indicated vs. Actual SCC Depth for Team MH.- RPC

0-

'a

(D

C:

w~

80

60

40

20

0

0 20 40 60

Actual Crack Depth, %

80 100

FIGU[E 3.64. EC Indicated vs. Actual SCC Depth for Team MO - Array Coil

3.111
2.5

2.0

4-

c 1.5

•1.0

.0

w 0.5

0.0 -

0.0

FIGURE 3.65. E4

2.5-

2.0

1.5

1.0-

-0 1.5

0(U

o(o

0.0

0.5 1.0 1.5 2.0

Actual Crack Length, in.

2.5

C Indicated vs. Actual SCC Length for Team MD - Bobbin

0.5 1.0 1.5

Actual Crack Length, in.

2.0 2.5

FIGUIE 3.66. EC Indicated vs. Actual SCC Length for Team MF - Bobbin

3.112
15)

C)

I.2.

0

2.5

2.0

1.5

1.0

0.5

0.01-

0.0 0.5 1.0 1.5 2.0

Actual Crack Length, in.

2.5

FIGURE 3.67. EC Indicated vs. Actual SCC Length for Team MH- RPC

2.5

C

U)

-j

0

.2

0

C

a)

0

0-o

uw

0.0 0.5 1.0 1.5 2.0

Actual Crack Length, in.

2.5

FIGURE 3.68. EC Indicated vs. Actual SCC Length for Team MO - Array Coil

3.113
cautioned that this round robin was obviously not designed to be a definitive

study, but rather to indicate trends. Too few specimens were employed to

permit firm conclusions about the capability of any given team to detect and

size SCC, or allow team-to-team comparisons.

3.8 COOPERATIVE RESEARCH NOT FUNDED BY PROJECT

The service-degraded Surry steam generator, located in a research facility,

presented a unique opportunity for testing and development of new technology

for application in the field. While the SGGP did not have program goals related

to the development of new technology, the Project made facilities available, on

a cost-reimbursement basis, for research, development, and demonstration

activities using the Surry generator. Such additional activities, subject to

SGGP sponsor acceptance, were only undertaken if they did not interfere with

the progress or goals of the SGGP. The results were communicated to the SGGP

sponsors, although specific details concerning the nature of the new technology

remained the property of the companies conducting the work. An advantage of

using the Surry generator and the SGEF was an ability to conduct the work in a

semi-realistic environment, including radiation exposure and service-degraded

materials.

As mentioned previously, several SGGP subcontractors entered into cost

sharing arrangements, allowing demonstration of their technology during use on

the Surry generator to perform needed tasks. This included the two dilute

chemical techniques demonstrated during the channel-head decontamination task,

and the development of automated equipment for tube plug removals. Most nonProject developmental activities involved the demonstration of repair techniques

or maintenance procedures which had been previously developed and tested in a

laboratory environment.

The maintenance procedures demonstrated included techniques to internally

sleeve damaged steam generator tubes, by brazing or welding an ID insert into

the affected tube region. Replacement of damaged tubing below the first support

plate was demonstrated. This involved internally cutting the tube above the

damaged region and then removing the damaged tube section through the tube sheet

by using normal tube-pulling techniques. The hole in the tube sheet was then

cleaned out and the cut stub of the undamaged portion of tube was cleaned and a

smooth end-finish machined. A matched section of new tubing was next welded in

the generator onto the old tubing. This required maintaining an inert atmosphere

in the generator. Final operations were expanding the new tube section into

the tube sheet and welding the end of the tube to the tube sheet. Quality

control consisted of NDE inspections using ultrasonics, along with visual

inspection.

Another maintenance operation demonstrated was the replacement of corrosion

damaged antivibration bars. This involved cutting the old bars, extracting

them from the tube bundle, then replacing the AVBs with new ones designed to

assist the replacement activity. The Surry demonstration helped to determine

the forces necessary to remove partially corroded bars, and allowed evaluation

of the effect of adjacent tubes of pulling the AVBs out. The demonstration

also permitted testing and refinement of the AVB replacement procedure and an

evaluatiun of the time and exposure required to perform this operation.

3.114
4.0 CONCLUSIONS

This section reviews the significant results and conclusions from all

phases of the SGTIP. The conclusions presented here served as the bases for

recommendations for improved'inservice inspection procedures and methods for

maintenance of steam generator tube integrity presented in Section 5.0.

4.1 TUBE INTEGRITY-

.Burst testing of steam generator tubes with artificial and service-induced

degradation'indicated that'tube failure was governed-by yielding of the

undegraded ligament. The basis for this conclusion is the fact that the burst

pressure of degraded steam generator tubes is, in most instances,

conservatively predicted by the axial flaw tables given in Section XI,

Subsection IWB-3640 of the ASME Code. This tables were derived from net

section collapse theory and validated'through tests on flawed stainless steel

piping. Since it is generally accepted that stainless steel pipes fail by

net section collapse, it seems reasonable to postu.late that Inconel 600 steam

generator tubes would'behave simi~larly.'

The burst test results also indicated that the most severe flaw types are

axial cracks-and uniform thinning. These flaws caused a larger decrease in

burst strength compared to' elliptical; wastage of similar dimensions.

Because of the high mechanical strength, ductility, and inherent

overdesign of steam generatbr'tubing, long "(>'one tube diameter) axial flaws

up to 80% through-wall did not rupture at pressures that would be experienced

during a main-steam-line-break' (MSLB) accident.,: Burst pressures several times

the normal operating or- main-steamii-line-break pressure differential were

measured for tubes with short axial" artificial flaws (i.e., <0.25 in.). This

result was validated by testing of tube segments with pitting/wastage type

defects removed from the Surry, generator. The Phase I empirical relationships

predicted the burst strength of service-degraded tubes with pitting/wastage

degradation to within'lO% of measured values. Even through-wall flaws of

this length would not burst under MSLB loading conditions. The test results

also showed that the tube burst strength.did not decrease significantly with

increasing axial flaw length for lengths greater than about one tube diameter.

For flaw lengths greater than-this level, the burstipressure was controlled

by the flaw depth.

All of the collapse pressures for'specimens with machined flaws were

higher than would be experienced during a loss-of-coolant-accident, which is

the worst credible collapse mode accident condition. EDM slots of any size had

very little effect on the collapse pressure. For all heats and sizes of

tubing with EDM slots up to 80% through-wall, the maximum decrease in collapse

pressure was 23%. The lowest collapse pressure of any EDM slot specimen was

3.6 times higher than the most severe pressure attainable during a LOCA. For

uniform-thinning and elliptical-wastage specimens, the collapse pressure

decreased with increasing flaw length and depth. Even for tubes with 75%

uniform thinning or elliptical wastage, the worst collapse pressure measured

was still -ppreciably higher than the expected LOCA collapse pressure.

4.1
Empirical equations were developed from the Phase I and Phase II pressure

test data by least squares regression techniques. The equations may be used

to predict the percent reduction in the strength of flawed steam generator

tubes of any geometry. Only the unflawed burst or collapse strength must be

known. The equations relate flaw size and orientation to remaining tube

strength and were validated by burst testing service-degraded tubes removed

from the Surry generator. The ASME IWB-3640 axial flaw tables bound most of

the SGTIP data and suggests that failure of steam generator tubes is governed

by net section collapse of the undefected ligament. Application of all the

IWB-3640 tables for axial and circumferential flaws to steam generator tubing

would extend the range of loading conditions and flaw geometries that could

be evaluated. For example, this would permit bending loading conditions to

be evaluated.

The main conclusion of the evaluation of leak rate data is that leak rates

measured from laboratory tests can be highly variable. These leak rates can

be influenced by "random" variables that are not addressed in the predictive

models. Measured leak rates are sometimes as much as a factor of ten less than

predicted rates. This suggests that conservatism should be applied in

calculations used to establish leak detection limits for detection systems.

It is concluded that the prediction of leak rates under ideal conditions

of crack opening behavior is a relatively straightforward calculation.

Differences in predictions from independently developed models were small

compared with the variability in the test data. Continued efforts to refine

the predictive models do not appear warranted, since it is likely that more

refined models would still not adequately treat the factors that give rise to

the variability in measured leak rates~in test specimens. The recommended

approach would be to recognize the uncertainties in predictions of leak rates,

and to use conservative margins in the applications of calculated leak rates.

The available leak rate data should also be viewed in a positive

perspective, since in most (but not all) tests the measured leak rates

approached or even exceeded the predicted rates. This clearly shows the value

of leak detection systems as a means to detect tube degradation prior to

failure of tube by rupture. Operating histories of reactors include numerous

cases of successful leak detection. The data~base addressed in the present

study'included many tests of tubes at pressures corresponding to a MSLB

accident. It is encouraging to note that none of this limited sample of tubes

experienced what could be called a rupture, whereas in most cases the measured

leak rates at normal operating pressures would have been detectable and

exceeded plant technical specification limits.

Deep axial SCC were found in tubes removed from.the Surry generator with

calculated strains as low as 10%. This indicates a strain-based plugging

criterion of 17% to 25% may be too high and that 10% is more appropriate.

4.2
4.2 NDE RELIABILITY

Laboratory measurements (under non-blind test conditions) of EC sizing

capability of precision-machined flaws indicated that sizing accuracy and

precision depended on the degradation volume. The depths of flaws such as

EDM notches and elliptical wastage were overestimated when the depth of

penetration was between 20% and 35%. On the other hand, EC underestimated " the depth for these flaw types when the depth of penetration was more than

50%. In addition, the degree of undersizing increased with increasing flaw

depth. The magnitude of the undersizing and scatter in the EC data increased

as the volume of metal removed decreased. For degradation, such as uniform

thinning, EC estimated depths were very precise, with a trend towards

oversizing at all flaw depths. Estimates of length 'sizing capability for

EDM slots and uniform-thinning flaws indicated that sizing accuracy was

similar, but that sizing precision depended on the volume of metal removed.

As with the depth sizing results, the length sizing precision was better for

flaw morphologies in which more of the tube wall was removed.

The POD for machined flaws also depended on volume of metal removed.

The POD increased with increasing volume of metal removed for flaws in the

20% to 35% depth range. EDM slots between 20% and 31% had a POD of 0.13,

whereas for elliptical-wastage and uniform-thinning specimens the POD was 0.8

and 1.0, respectively. When the flaw depth was greater than 50%, the POD was

1.0 for all flaw types.

The single-frequency EC results for tubes with chemically-induced

degradation were similar to the trends observed for specimens with machined

degradation. As with the EC sizing capability studies conducted on precisionmachined flaws, these tests were performed under non-blind laboratory

conditions. Single-frequency EC measurements of laboratory SCC depth indicated

that this type of degradation was, on the average, undersized. The mean error

was found to be -1.3% of wall. Data scatter for this defect type was quite

high, with a standard deviation of 17.1. (For additional conclusions regarding

EC sizing reliability of SCC, see discussion below on SCC mini-round robin

results.) Specimens with chemically-produced elliptical wastage also tended

to be undersized in depth (-5.3% of wall mean error), but data scatter was

much less than for the SCC-degraded tubes, with a standard deviation of only

3.8. Similar to the Phase I results, the tubes with chemically-produced

uniform thinning were conservatively oversized, with a mean error of about

9%. The standard deviation for this flaw type was comparable to the

elliptical-wastage specimens at about 4.0.

The single-frequency EC results discussed above should be considered

only as a rough indicator of EC capability, since 1) the tests were conducted

under non-blind laboratory conditions and 2) the test specimens all had

idealized flaws and no complicating factors to simulate real-world conditions,

such as copper deposits, denting, support plates, and antivibration bars. In

addition, the tests were performed so that inspectors were aware of the type

of degradation being examined. The Surry investigation provided a unique

opportunity to gather NDE reliability information on tubes with service-induced

degradation and extraneous effects under simulated field conditions using

multifreque~ncy EC equipment. Thus, the information developed more closely

4.3
represents the reliability of actual inservice EC inspections rather than a

measure of the capability of the equipment to detect and size degradation in

ideal. situations. Significant findings from the numerous round robins

performed on the Surry generator are summarized below.

Metallurgical and visual examination of specimens from all regions of the

Surry 2A steam generator identified a variety of defect types at specific

locations. Pitting and wastage were the predominant defects present. These

defects were located in the sludge pile region above the TTS, within the

crevice region between the tube and support plates, and, to a lesser extent,

at antivibration bar contact areas. The severity of the pitting/wastage

degradation was generally <20% through-wall except for specimens from the

sludge pile region above the TTS where the wall-loss ranged up to 87%.

Sufficient numbers of specimens with other types of degradation were not

found, and thus the following conclusions regarding the reliability of EC

inspections in detecting and sizing flaws relates primarily to pitting/wastage

type degradation.

The probability of detecting service-induced flaws increased with

increasing wall-loss and approached 0.9 for pitting/wastage degradation more

than 40% through-wall. An automated data screening technique employed by one

team that inspected the Surry generator appeared to improve the probability for

detecting pitting/wastage degradation over conventional EC analysis methods.

Eddy-current estimates of the through-wall depth of service-induced

pitting/wastage degradation showed wide variations between teams for the same

flaw indication and also between specimens with similar wall losses. The

team-to-team variations from a given specimen appeared to result from

differences in analysis procedures and analyst interpretation. This

team-to-team variance could not be significantly reduced by using consistent

analysis procedures. The specimen-to-specimen variations appeared to be

related to the complex flaw morphology which produces complex EC signals.

Conventional EC tended to underestimate the depth of service-induced

pitting/wastage degradation. Sizing accuracy of this type of degradation

decreased as the complexity of its morphology and through-wall depth

increased. Improved sizing accuracy on service-induced degradation was

obtained for one team that employed special frequency mixes to enhance the

signal-to-noise ratio by suppression of signals due to denting, copper

deposits, and support plates. Also, ultrasonic and rotating EC probes were

successfully used to augment conventional EC/bobbin-coil data and provide an

improved estimate of pitting/wastage depth.

Denting at tube support plate intersections in the Surry Generator

interfered with the EC inspections and made flaw detection and sizing nearly

impossible at these locations.

Copper-rich deposits found on the OD surfaces of tubes removed from the

Surry Generator produced EC signals, which resulted in false calls being made

in the U-bend and support-plate regions of the generator. Flaw-like EC signals

were produced at regions where there was an interruption in the Cu-rich

deposits.

4.4
To supplement the NDE vallidation data obtained from the SGGP and obtain

a more realistic assessment of the reliability of NDE techniques to detect and

size SCC in steam generator tubes under simulated service conditions, an

additional mini-round robin was performed. The results from the mini-round

robin using a limited number of tubes from the Phase II specimen inventory,

with laboratory-produced SCC, indicated that the average POD by conventional

multifrequency EC using bobbin-coil probes was only 0.5. Alternate inspection

techniques, using rotating pancake-coil..,. array-coil and segmented bobbin-coil

probes also showed a low POD for SCC, with an average of 0.63. The reliability

of both conventional bobbin-coil and alternate EC techniques to size

laboratory-produced SCC length and depth was not very accurate or precise.

The reader should be cautioned.that this round robin was not designed to be a

comprehensive study, but rather to indicate trends.

4.3 EVALUATION OF SAMPLING PLANS FOR ISI

The analytical evaluation of 100% inspection indicated that, on the

average, most teams that.inspected the steam generator would identify only

about 65% of the.defective.tubes,(defined as tubes with >75% through-wall

degradation). In contrast, the best performing team could be expected to

identify about 95% of the defective tubes.

The results of the analytical evaluation also demonstrated that if the

clustering assumption holds, and if the POD for degraded but not defective

tubes was at least 0.7, then the sequential sampling plan with 40% systematic

sampling at the initial.stage was nearly as effective as 100% inspection for

detecting and plugging defective tubes. This was true for any PEL value.

However, the 20% systematic sampling at the initial stage was significantly

inferior to both 100% inspection and the 40% systematic sampling plan.

The results of the Monte Carlo simulation analysis supported the

conclusions reached from the analytical evaluation and provided some valuable

additional insights. When the isolated clustering assumption discussed in

Section 3.6.4 was approximately valid, the 40% sequential scheme tended to be

nearly as effective as 100% inspection, but required substantially less

inspection. However, when the defective tubes were in one large cluster with

degraded tubes, all three sampling plans were equally effective for detecting

and plugging defective tubes.

Improving the sizing capability to that demonstrated by Team V resulted

in improved effectiveness of all three inspection plans. Also, with the

improved sizing capability, increasing the EC plugging limit from 40% to 50%

had little impact on the effectiveness of each sampling plan for plugging

defective tubes, but it minimized the probability for plugging nondefective

tubes. Whereas, if the "Average U.S. Team" sizing capability was assumed,

increasing the plugging limit from 40% to 50% significantly decreased the

effectiveness of all sampling plans for plugging defective tubes.

Improving the POD enhances the effectiveness of all inspection plans,

and the best performance was achieved when the better POD curve and the better

sizing model were assumed. For the POD curves considered, flaws with true

4.5
sizes less than 20% through-wall did not significantly improve the

effectiveness of the sampling plans because these flaws had a very low POD.

4.4 STEAM GENERATOR PREPARATORY TASKS AND INITIAL INSPECTIONS

Secondary-side examinations conducted on the Surry Generator were useful

for qualitatively identifying conditions that could lead to loss of steam

generator integrity. These examinations did not provide quantitative tube

degradation information usable for assessing~remaining tube integrity.

Dilute chemical decontamination such as by the LOMI or CAN-DECON

techniques (modified for PWRs) provided decontamination factors up to ten in

the channel-head region on the Surry Generator. No significant corrosion

effects were observed from these decontamination processes. In addition, the

volume of liquid waste could be minimized through the use of ion exchange

resins.

Radiation monitoring of channel-head workers during the tube unplugging

task indicated that exposure to the eyes was the limiting exposure. Thus,

badging of the forehead and chest was determined to be the most effective for

ensuring adequate personnel dosimetry.

4.6
5.0 RECOMMENDATIONSJFOR IMPROVED ISI AND TUBE PLUGGING LIMITS

This section presents recommendations for improved ISI and tube plugging

limits based on the research results of the SGTIP/SGGP.

5.1 RECOMMENDATIONS FOR IMPROVED ISI

5.1.1 Full-Length Tube Inspection

It is recommended that tubes should be inspected full length, from tube

end to tube end. This recommendation is based on observation of significant

degradation on the cold-leg portion of Surry tubes and occurrences of tube

rupture events in this region of the generator.

5.1.2 Screening of Data Prior to Analysis

All data should be screened prior to analysis. An automated data

screening system may be used, but it should be established via performance

demonstrations that it is capable of detecting all flaws including small-volume

degradation such as IGA, SCC,' and fatigue cracking. -The basis for this

recommendation is in the superior POD performance of one of the teams that

inspected the Surry Generator. By employing an automated data screening

system to filter the vast number of EC signals, this:team had the highest

overall POD performance.

5.1.3 Independent Data Analysis

As an additional aid towards improved flaw detection and characterization,

it is recommended that an independent review of the EC inspection data be

performed. Although independent reviews of the data were not done during the

inspections of the Surry 2A Steam Generator, there were several instances in

which a review of the data could have helped avoid missed flaws. There were

several instances noted during the validation in which one team missed a flaw

that other teams detected. In addition, the team-to-team variation in flaw

sizing was substantial for specimens with similar wall losses. It is

recommended that accord be reached between the principal and independent

analyst regarding the nature and categorization of the eddy-current data. If

accord cannot be reached, then conservative practice would be to reinspect

the tube with supplemental inspection methods or-plug/repair the tube.

5.1.4 Sampling Plan

Based on a statistical evaluation and Monte Carlo studies of different

flaw distributions and different inspection plans and using the Surry roundrobin results, the recommended minimum sampling scheme should be a systematic

sequential sample with 40%- sampling at the first stage. During each inservice

inspection, a sample of tubes should be selected according to a 40% systematic

sampling grid. An example of a 40% systematic sampling grid pattern is

illustrated in Figures 5.1 to 5.3, where sampled tubes are denoted by X.

Each tube in the inspection sample plan should be inspected over its full

length frori tube end to tube end. For each tube in the sample with an

5.1
ABCDEFGHIJKLMNOPQRSTUVWXY

1 XXOOOXXOOOXXOOOXXOOOXXO00

2 OOXXOOOXXOOOXXOOOXXOOOXXO

3 XOOOXXOOOXXOOOXXOOOXXOOOX

4 OXXOOOXXOOOXXOOOXXOOOXXOO

5 OOOXXOOOXXOOOXXOOOXXOOOXX

6 XXOOOXXOOOXXOOOXXOOOXXOO0

7 OOXXOOOXXOOOXXOOOXXOOOXXO

8 XOOOXXOOOXXOOOXXOOOXXOOOX

9 OXXOOOXXOOOXXOOOXXOOOXX00

10 O00XXOOOXXOOOXXOOOXXOOOXX

FIGURE 5.1. 40% Systematic Sampling

Grid Pattern for First

Inservice Inspection

ABCDEFGHIJKLMNOPQRSTUVWXY

I OOXXOOXXOOOXOOOXXOO00XX0O

2 XOOOXXOOOXXOOOXXOOOXXOOOX

3 OXXOOOXXOOOXXOOOXXOOOXXOO

4 OO0XXOOOXXOOOXXOOOXXOOOXX

5 XXOOOXXOOOXXOOOXXOOOXXOOO

6 OOXXOOOXXOOOXXOOOXXOOOXXO

7 XOOOXXOOOXXOOOXXOOOXXOOOX

8 0XXOOOXXOOOXXOOOXXOOOXXO0

9 O00XXOOOXXOOOXXOOOXXOOOXX

10 XXOOOXXOOOXXOOXXOOOXXOO00

FIGURE 5.2. 40% Systematic Sampling

Grid Pattern for Second

Inservice Inspection

ABCDEFGHIJKLMNOPQRSTUVWXY

1 XO00XXOOOXXOOOXXOOOXXOOOX

20OXXOOOXXOOOXXOOOXXOOOXXOO

3 OO0XXOOOXXOOOXXOOOXXOOOXX

4 XXOOOXXOOOXXOOOXXOOOXXO0O

5 00XXOOOXXOOOXXOOOXXOOOXXO

6 XOOOXXOOOXXOOOXXOOOXXOOOX

7 OXXOOOXXOOOXXOOOXXOOOXXOO

8 OOOXXOOOXXOOOXXOOOXXOOOXX

9 XXOOOXXO00XXOOOXXOOOXXO00

10 0OXX00OXXO00XXOOOXXOOOXXO

FIGURE 5.3. 40% Systematic Sampling

Grid Pattern for Third

Inservice Inspection

5.2
eddy-current indication due to degradation (of any size), further inspection

should be performed by testing tubes immediately surrounding the tube with

the eddy-current degradation indication. This additional inspection should

continue until two-tube-wide "buffer zone" is observed, which is composed of

tubes with no eddy-current degradation indications and which completely

surrounds the tube(s) with eddy-current degradation indications. In

implementing this sampling scheme, the initial systematic sampling grid should

be "shifted" at each inservice inspection so that every tube in a generator

will be inspected at least once during the course of three consecutive

inservice inspections and at least twice during the course of five inservice

inspections. In addition, since inspection results from one steam generator

are not always reliable indicators of the condition of other steam generators

in the plant, it is also recommended that every steam generator be inspected

at each inservice inspection.

5.1.5 Application of Frequency Mixes for Noise and Foreign Deposit

Suppression

Analysis of the Surry round-robin data indicated that special frequency

mixes designed to suppress the effects of internal support structure, denting,

and copper deposits was successfully used by one team. This team exhibited the

highest POD and flaw sizing accuracy of all the teams inspecting the Surry

Generator. The use of frequency mixes specially developed for the Surry

Generator improved the signal-to-noise ratio over that attainable with

conventional inspection procedures, and provided enhanced flaw detection and

sizing capability. The teams participating in the Base-line and DAARR round

robins used multifrequency eddy-current inspection equipment, but during data

evaluation only a single frequency (400 kHz) tended to be used (Doctor et al.

1986b). Thus, much of the capability of multifrequency inspection was not

utilized in evaluating the inspection results.

5.1.6 Development of Performance Demonstrations for Personnel, Equipment

and Procedures

The foregoing research results indicate that existing requirements for

qualifying personnel, equipment and procedures for eddy-current inspection of

nuclear steam generators may not be adequate to ensure reliable inspection

performance. In order to obtain effective inservice inspection, an NDE system

qualification covering the personnel, equipment and inspection procedure is

recommended. In this context, personnel qualification consists of training,

qualification and certification. The inspection personnel, equipment and the

inspection procedure should be qualified in combination through performance demonstrations. One possible performance demonstration program might consist

of two components: a generic qualification, and a plant-specific

qualification. A generic performance demonstration is needed to ensure that

inservice inspections are effective at detecting any type of degradation that

may occur in any steam generator. On the other hand, because different plants

often exhibit different degradation mechanisms and conditions, it is important

that inspection personnel should also be qualified to perform inspections on

that particular plant.

5.3
5.1.7 Improved Techniques for Crack Detection

Results from the SCC mini-round robin suggest that crack detection and

sizing may be a significant problem. These types of flaws had the lowest POD

and among the worst sizing accuracy for any of the types of degradation

investigated. It was also concluded that along with uniform-thinning

degradation, this class of defects caused the largest decrease in tube burst

strength. Further, as noted from the survey of laboratory data summarized in

Appendix A, crack-type flaws can grow at very high rates, potentially from

initiation to leakage in less than one operating cycle. In consideration of

these concerns, it is recommended that if cracking degradation is known or

suspected, then NDE techniques optimized for crack detection be employed.

5.2 RECOMMENDATIONS FOR IMPROVED TUBE PLUGGING LIMITS

5.2.1 Through-Wall Flaws

It is recommended when the plant is shut down that all tubes with

through-wall flaws discovered during operation, or found during inservice

inspection, should be plugged or repaired. This recommendation stems from the

evaluation of the publicly available leak-rate data. This evaluation showed

that through-wall flaws, in some cases, leaked significantly less than

predicted from model calculations. The lower than expected leakage may have

been caused by the complex interaction of "random" variables such as

1) residual stresses which hold the crack closed, 2) crack roughness leading

to increased fluid flow resistance, or 3) crack plugging by impurity or

corrosion product build-up on fracture faces. Any or all of these effects

could have influenced the measured leakage and might also be operative for

inservice cracks. Given the unpredictability of leak rates from cracks in

steam generator tubes, the conservative approach to management of known

through-wall flaws would be to plug or repair the tube.

5.2.2 Disposition of Crack-Type Flaw Indications

Based on the above discussion and a cursory literature search to document

flaw growth rates (see Appendix A), it is recommended that all tubes with

indications of cracking should be plugged or repaired. It is recognized that

under some conditions the return to service of a cracked tube may be justified.

For example, there are many examples of tube cracking within the tubesheet

where the tube is restrained by frictional forces caused by tube expansion.

Thus, if compelling engineering evidence can be developed to demonstrate that

reliable NDE techniques and procedures are available for crack detection and

sizing, and that reliable and conservative data on expected crack growth until

the next inspection is also available, then it is recommended that case-bycase exceptions to plugging of all tubes with crack-type indications be

considered.

5.2.3 Flaw Characterization and Consideration of Inspection Uncertainties

Based on the pressure test results and a review of the flaw growth rate

literature, it is evident that degradation type can significantly affect proper

flaw evaluition. In other words, flaw type (i.e., cracks, pits, wastage,

5.4
erosion-corrosion, etc.), geometry and orientation were found to be the most

important variables in determining the burst or collapse pressure of a degraded

tube. In addition, flaw growth rates depend largely on the type of flaw under

consideration. In order to perform the most accurate flaw evaluation possible,

it is desirable to obtain specific knowledge of the type of degradation

afflicting a particular tube. Thus, it is recommended that other types of

information such as tube pulls and secondary-side inspections be used to

supplement conventional eddy-current inspection results.to permit improved

characterization of tube degradation. The locations of indications relative

to support plates, tube sheets, and U-bends can also provide useful

information for establishing the likely modes of degradation. Past service

histories for the particular steam generator and for other generators of

similar design and operating conditions may also be useful for establishing

the degradation type.

The NDE flaw-sizing error is also an important consideration in

determining an appropriate tube plugging limit. An accurate estimate of the

degradation size in terms of depth, length, orientation and proximity to other

degraded regions is needed in order to arrive at a realistic estimate of

remaining tube integrity. It is recommended that the degraded region should

be characterized by bounding dimensions so that it can be described by

parameters of depth, orientation, and a length dimension as measured along

the axial and circumferential directions of the tube.

A considerable body of EC sizing error data was developed during the

SGTIP on typical steam generator tubes with both natural and artificial flaws.

This extensive information base indicates that the EC sizing error depends

strongly on the type of degradation, the inspection equipment and procedure,

and the presence of extraneous conditions such as surface deposits. Even

though the SGTIP data provides useful estimates of the EC sizing error in the

absence of more definitive information, it is recommended that specific

allowances for flaw sizing uncertainty be developed from NDE performance

demonstrations (see discussion in 5.1.6 above). Such performance

demonstrations should be based on a statistically significant number of samples

for the relevant type of degradation and NDE method employed. Since the SGTIP

pressure test results show that relatively deep but short degraded regions

have little effect on burst pressures, reductions in uncertainties in length

measurements should be given equal priority to improved accuracy of depth

measurements.

5.2.4 Methods for Flaw Evaluation

An extensive amount of information on the effect of flaw type and severity

on remaining tube integrity was developed during the SGTIP. Empirical

equations were developed relating the extent of degradation to remaining tube

integrity. It is recommended that evaluations of the effect of degradation

on tube integrity should be based on the empirical equations of burst tests

of tubes with artificial and/or service-induced degradation. Since the

empirical data on tube failure are limited to pressure loadings, the effects

of bending loads need to be addressed by predictive models. If ductile

fracture is assumed, then the net section collapse of the remaining

cross section of the tube as presented in Section XI, Subsection IWB-3640, of

5.5
the ASME Code, 1986 Edition, is recommended as a suitable model for evaluating

these conditions. The bounding trends of tube burst tests are summarized in

Figure 5.4 for degradation geometries and loads such that the circumferential

stress governs predictions of tube integrity. This figure is an example of

the data recommended for predicting the integrity of degraded tubes. The

curves indicate percentage reductions in allowable pressure as a function of

both degradation depth and length. Figure 5.5 gives curves that are

recommended for evaluating circumferentially-oriented degradation for which

the critical stress is in the axial direction. These curves are purely for

pressure loading. Limited tube burst tests served to confirm the trends of

these curves. The technical basis for generating these curves was the IWB-3640

tables for net section collapse of the remaining wall cross section of the

degraded tube. Extension of these results to account for bending loads could

be made using the net section collapse methodology.

1.0

0

a0.8

0.6

0.4

0.2

0.0

0 5 10 15 20

L/4R't

FIGURE 5.4. Burst Parameter Prediction Curves for Axially Oriented

EDM Notches in Steam Generator Tubes. L = Degradation

Length, a = Degradation Depth, R = Tube Inner Radius,

and t = Tube Wall Thickness

5.6
.06

0.2"

Z

0.0

0 510 1520

L/4R*t

FIGURE 5.5. B ur .st Parameter Prediction Curves for Circumferentially

Oriented Cracks (based on IWB-3640). L = Degradation

Length, a = Degradation Depth, R =Tube Inner Radius,

and t = Tube Wall Thickness

5.2.5 Repair Versus Plugging

As an alternative to plugging of defective tubes, it is recommended that repair be considered a viable option for maintaining steam generator

integrity. The repair should meet the same performance requirements (i.e.,

ASME Code) as the original tube, including structural integrity and leak

tightness. The repaired tube should be inspectable both for the repaired

area and for continued inspection of the unrepaired length of the tube.

Activities in the QC/QA area should assure that all defective tubes have

actually been repaired. Accurate documentation through photographs, video

tapes, or other methods should be utilized.

5.2.6 Ptugging Limit for Allowable Denting

Operating experience with steam generators at a number of plants has

resulted in severe build-up of corrosion products in the tube-to-tube

support-plite crevice region, resulting in denting of the tube. While this

5.7
reduction in tube diameter does not by itself constitute a problem to tube

integrity, it has resulted in initiation and growth of stress corrosion cracks

that may lead to leakage and rupture. In addition, the reduction in tube

diameter could prevent the passage of appropriately-sized NDE probes, making

inservice inspections unreliable or impossible.

It is' recommended that plugging limits for dented tubes be based on either

reduction in tube diameter or on peak strain at the dented location of the

tube. Peak strains can be calculated from measurements of the profile of the

dented tube as demonstrated during-the PROFILE-360 examination of selected

tubes from the Surry Generator.

Based on the work, an appropriate strain limit for plugging of dented

tubes would be 10%. It is recognized that the susceptibility of tubes to SCC

is governed by factors such as strain rate, material microstructure, and plant

water chemistry. Thus, strain-based plugging limits lower than 10% may be

needed for particular plants. Development of an alternate plugging limit

should seek to preclude crack initiation and limit denting severity such that

effective tube inspection is possible.

5.2.7 Flaw Growth Allowance

One of the considerations in determining tube plugging limits is an

allowance for degradation growth until the next scheduled inservice inspection.

The rationale for including this allowance is to 1) ensure that no degradation

grows through-wall, and 2) that all degradation remains at an acceptable size

consistent with required safety margins.

Based on the cursory literature review given in Appendix A, it is evident

that predictions of degradation growth are subjectto considerable

uncertainty, and, therefore, it is not possible to recommend precise methods

or supporting data for this purpose. A practical method for determining this

allowance that would be relevant to specific plants is the application of

historical service experience and results from inservice inspections. In

addition, observed growth rates for plants with similar operating histories

should be considered. Since the laboratory data indicates there are vast

differences in growth rates for different degradation mechanisms, the growth

for the specific mechanism should be determined. If knowledge of the

degradation mechanism is lacking, then growth rates for the most rapidly

propagating degradation mechanism should be applied.

Growth allowances based on observed plant-specific rates should be subject

to conservative assumptions about future service conditions and to confirmation

of degradation extent during scheduled inspections. Growth rates less than

projections based on the prior.observed rates should be used only if

substantial reductions in the severity of the operating environment can be

assured. If methods other than linear projections of observed growth rates,

are used, then predictions of these other methods should be checked for

consistency with the observed rates of growth.

5.8
6.0 REFERENCES

Aitkin, M. 1981. "A Note on the Regression Analysis of Censored Data."

Technometrics 23(2):161-163.

Allen, R. P., R. L. Clark, and W. D. Reece. 1984. Steam Generator Group

Project Task 6 - Channel Head Decontamination. NUREG/CR-3841, PNL-4712,

Pacific Northwest Laboratory, Richland, Washington.

Alzheimer, J. M., R. A. Clark, C. J. Morris, and M. Vagins. 1979. Steam

Generator Tube Integrity Program - Phase I Report. NUREG/CR-0718,

PNL-2937, Pacific Northwest Laboratory, Richland, Washington.

American Society of Mechanical Engineer's Boiler and Pressure Vessel Code.

Section XI Rules for In-Service Inspection of Nuclear Power Plant Components. 1986 Edition. ASME, New York, New York.

Angwin, M. J. et al. 1985. Steam Generator Reference Book. Electric Power

Research Institute, Palo Alto, California.

Aspden, R. G. 1986. "Background Reporting and IGA Screening in Static Test

Devices." In Proceedings: 1984 Workshop on Secondary Side Stress

Corrosion Cracking and Intergranular Corrosion of PWR Steam Generator

Tubing. EPRI NP-4478, Electric Power Research Institute, Palo Alto,

California.

Bandy, R. and D. Van Rooyan. 1984. "Stress Corrosion Cracking of Inconel

Alloy 600 in High Temperature Water - An Update." Corrosion

40(8):425-430.

Baum, A. J., C. Dangelo, P. J. Kuchirka, N. Singleton, and M. J. Sredzienski.

1987. Steam Generator Cold Leg Thinning in Operating Plants. EPRI

NP-5140, Electric Power Research Institute, Palo Alto, California.

Berge, J. P. 1986. "Questions on U-Bend Cracking." In Proceedings: 1984

Workshop on Secondary Side Stress Corrosion Cracking and Intergranular

Corrosion of PWR Steam Generator Tubing, EPRI NP-4478, Electric Power

Research Institute, Palo Alto, California.

Berge, J. P. 1987. "Overview of Primary Side Stress Corrosion Cracking SG

Tubes of Electricite de France." In Proceedings: 1985 Workshop on

Primary-Side Corrosion Cracking of PWR Steam Generator Tubing,

EPRI NP-5158, Electric Power Research Institute, Palo Alto, California.

Bickford, R. L., R. A. Clark, P. G. Doctor, H. E. Kjarmo, and C. A. LoPresti.

1984. Eddy-Current Round Robin Test on Laboratory Produced Intergranular

Stress Corrosion Cracked Inconel Steam Generator Tubes. NUREG/CR-3561,

PNL-4695, Pacific Northwest Laboratory, Richland, Washington.

6.1
Bowen, W. M., P. G. Heasler, and R. B. White. 1989. Evaluation of Sampling

Plans for In-Service Inspection of Steam Generator Tubes: Part I.

NUREG/CR-5161, PNL-6462, Pacific Northwest Laboratory, Richland,

Washington.

Bradley, E. R., P. G. Doctor, R. H. Ferris, and J. A. Buchanan. 1988. Steam

Generator Group Project Task 13 Final Report - NDE Validation.

NUREG/CR-5185, PNL-6399, Pacific Northwest Laboratory, Richland,

Washington.

Brown, S. D. 1985. PWR Steam Generator Tubing Inspection Guidelines.

Electric Power Research Institute, Palo Alto, California.

Clark, R. A. and M. Lewis. 1984.

Report - 1982. NUREG/CR-3581, Richland, Washington.

Clark, R. A. and M. Lewis. 1985.

Report - 1984. NUREG/CR-4362,

Richland, Washington.

Steam Generator Group Project Annual

PNL-4693, Pacific Northwest Laboratory,

Steam Generator Group Project Annual

PNL-5417, Pacific Northwest Laboratory,

Davis, C. S., J. M. Thomas, S. W. Winder, and D. E. Allison. 1983. An

Engineering and Probabilistic Analysis of Tube Cracking Performance in

Once-Through Steam Generators. EPRI NP-3065, Electric Power Research

Institute, Palo Alto, California.

DeRosa, P., W. Rinne, H.

Steam Generator Tube,

Plus SSE Conditions.

Pennsylvania.

W. Massie,

Tube Sheet

WCAP-7832,

and P. Mitchell. 1973. Evaluation of

and Divider Plate Under Combined LOCA

Westinghouse Electric Corp., Pittsburgh,

Doctor, P. G., J. M. McIntyre, P.K. Alley, L. R. DesCamp, J. S. Littlefield,

C. A. LoPresti, and J. A. Buchanan. 1983. The In-Service History of

The Surry Unit 2A Steam Generator. PNL-4880, Pacific Northwest

Laboratory, Richland, Washington.

Doctor, P. G., A. S. Birks, R. H. Ferris, H. Harty, and G. E. Spanner.

Steam Generator Group Project Task 7 Final Report - Post Service

Baseline Eddy-Current Examination. NUREG/CR-5087, PNL-5940, Pacific

Northwest Laboratory, Richland, Washington.

1988a.

Doctor, P. G., H. Harty, R. H. Ferris, and A. S. Birks. 1989. Steam

Generator Group Project Task 9 Final Report - Nondestructive Evaluation

Round, Vol. 1: Description and Summary Data. NUREG/CR-4849, PNL-5868,

Pacific Northwest Laboratory, Richland, Washington.

Hall, J. F. 1984. "Pitting in the Laboratory at Combustion Engineering."

In Workshop Proceedings: Pitting in Steam Generator Tubing. EPRI

NP-3574-SR, Electric Power Research Institute, Palo Alto, California.

6.2
Hall, J. F., R. S. Frisk, K. E. Marugy, and A. S. O'Neill. 1987.

Investigation of Causes and Corrective Actions for Pitting in Steam

Generator Tubes: Prototypic Tests. EPRI NP-5248, Electric Power Research

Institute, Palo Alto, California.

Hermer, R. E. and C. R. Wolfe. 1986. "Alloy 600 Intergranular

Corrosion in Alkaline Environments." In Proceedings: 1984 Workshop

on Secondary Side Stress. Corrosion Cracking and Intergranular

Corrosion of PWR Steam Generator Tubing. EPRI NP-4478, Electric Power

Research Institute, Palo Alto, California.

Jacko, R. J. 1986. "IGSCC and IGA Behavior of Inconel 600." In

Proceedings: 1984 Workshop on Secondary Side Stress Corrosion

Cracking and Intergranular Corrosion of PWR Steam Generator Tubing.

EPRI NP-4478, Electric Power Research Institute, Palo Alto, California.

Kurtz, R. J., R. L. Bickford, R. A. Clark, C. J. Morris, F. A. Simonen

and K. R. Wheeler. 1988. Steam Generator Tube Integrity Program - Phase II Final Report. NUREG/CR-2336, PNL-4008, Pacific Northwest

Laboratory, Richland, Washington.

Kurtz, R. J., M. Lewis and R. A. Clark. 1987. Steam Generator Group

Project Annual Report - 1985. NUREG/CR-4848, PNL-5771, Pacific Northwest

Laboratory, Richland, Washington.

Miglin, B. P.. 1987. Production of Eddy-Current Standards for Caustic

Intergranular Corrosion. EPRI NP-5109, Electric Power Research Institute,

Palo Alto, California.

Newman, J. R., P. McIntyre, and S. M. Payne. 1987. "Effect of Thermal

Treatment and Impurity Segregation on the Kinetics of IGA in Inconel

600." Presented at EPRI Workshop on IGA and SCC Mechanisms, Electric Power

Research Institute, Palo Alto, California.

Partridge, M. J. 1987. "Effect of Temperature and Caustic on IGA and

IGSCC Corrosion Rates." Presented at EPRI Workshop on IGA and SCC

Mechanisms, Electric Power Research Institute, Palo Alto, California.

Pathania., R. S. and P. V. Balakrishnan. 1986. Correlation of Tube

Support Structure Studies. EPRI NP-4672, Electric Power Research

Institute, Palo Alto, California.

Powell, D. E. and J. F. Hall. 1987. "Experimental Measurements of Leak

Rates and Burst Pressures for Cracked Alloy 600 Tubes Under Steam Generator

Conditions." In Proceedings: 1985 Workshop on Primary-Side Corrosion

Cracking of PWR Steam Generator Tubing, EPRI NR-5158, Electric Power

Research Institute, Palo alto, California.

Reece, W. D. and R. Harty. 1983. Dosimeter Placement and Interpretation

for Determination of Dose to the Extremities and the Head. PNL-4967,

Pacific Northwest Laboratory, Richland, Washington.

6.3
Sawsochka, S. G.,

Correlation of

Power Research

K. D. Friedman, S. S. Choi, and J. K. Pyo. 1986.

Tube Support Corrosion Studies. EPRI NP-4818, Electric

Institute, Palo Alto, California.

Schwenk, E. B. and K. R. Wheeler. 1984. Steam Generator Group Project

Task 10 - Secondary Side Examination. NUREG/CR-3843, PNL-5033, Pacific

Northwest Laboratory, Richland, Washington.

Schwenk, E. B. 1987. Steam Generator Group Project Task 10 - Secondary

Side Examination. NUREG/CR-4850, PNL-6045, Pacific Northwest Laboratory,

Richland, Washington.

Simonen, F. A. et al. 1986. Reactor Pressure Vessel Failure Probability

Following Through-Wall Cracks Due to Pressurized Thermal Shock Events,

NUREG/CR-4483, Pacific Northwest Laboratory, Richland, Washington.

Sinclair, R. B. 1984. Secondary Side Photographic Techniques Used in

Characterization of Surry Steam Generator. NUREG/CR-3094, PNL-5053,

Pacific Northwest Laboratory, Richland, Washington.

Theus, G. J. 1981. "Stress Corrosion Cracking Tests of Alloy 600." In

Workshop Proceedings: U-Bend Tube Cracking in Steam Generators,

EPRI WS-80-136, Electric Power Research Institute, Palo Alto, California.

Wheeler, K. R. 1984. Initial Inspection of the Service Degraded Surry

2A Steam Generator, June 1982. PNL-5268, Pacific Northwest Laboratory,

Richland, Washington.

Wheeler, K. R., P. G. Doctor, L. K.

Generator Group Project Task 8 -

NUREG/CR-3842, PNL-4876, Pacific

Washington.

Fetrow, and M. Lewis. 1984. Steam

Selective Tube Unplugging.

Northwest Laboratory, Richland,

6.4
.APPENDIX A

COMPENDIUM OF DEFECT GROWTH RATES
APPENDIX A

COMPENDIUM OF DEFECT GROWTH RATES

1.0 INTRODUCTION

The information in this appendix summarizes much of the currently available

data on the growth rate of the common defects found in steam generator

tubing. This information was compiled to assist the determination of appropriate degradation allowances for evaluation of degraded steam generator

tubing. Most of the data was obtained from recently published research

reports and for the most part represents the results of laboratory investigations conducted under severe test conditions (e.g., high applied stresses and

aggressive environments). In some cases, field data was available for

comparison with the laboratory 'produced data.

2.0 THINNING (WASTAGE)

Very little data on the growth rate of wastage is available. With no plants

in the U.S. presently using phosphate secondary water chemistry, this problem

is not a major concern. Most recent occurrences of wastage have been observed at the periphery of the first or second tube support plate on the cold

leg side of the generator. Limited data on the progression of cold leg

thinning was reported by Baum et al. (1987). These data were obtained by

monitoring the change in the eddy-current measurements on a given tube during

successive inspections and dividing by the elapsed time. Corrosion rates

were based on data from tubes with indications having reported depths 20% or

greater in two successive inspections. The calculations were based on a

total of 489 tubes in 18 steam generators. An overall degradation rate of 3

mils/y was estimated from these data.

Other published data (Angwin et al. 1985) plotted in Figures A.1 and A.2 show

that the corrosion rate depends upon the Na/P04 molar ratio and temperature,

but maximum corrosion rates never exceeded about 5.5 mils/y.

Estimates were also obtained from evaluation of Surry metallographic data

presented by Bradley et al. (1988). Maximum wall penetration of pure wastage

was measured on tubes removed from both the hot and cold leg sides of the

Surry generator. Table A.1 gives the minimum, average, and maximum percent

wall-loss measured.

The data in Table A.1 can be used to estimate growth rates by dividing by the

duration the defect mechanism is operative. The Surry generator was in

service for slightly less than six years, from March 1973 to January 1979.

A.1
(/}

.n

e

0o

0

5

4

3

2

1

1.0 1.35 1.6 2.0 2.3 2.7 3.0

Na/P0 4 38712-170.1M

FIGURE A.1. Average Corrosion Penetrations of

Phosphate Solutions at 615°F as a

Ratio (Angwin et al. 1985).

Alloy 600 in Deaerated Sodium

Function of the Na/P04 Molar

Table A.1.

Leq

CL**

HL

HL*

HL**

Maximum Percent Wall-Loss

Degraded Surry Tubes

Min

4

12

1220

Ave

11

30

22

35

for Wastage

Max

24

52

32

52

*Tubes never plugged during service

**Tubes plugged during service

A.2
(MC)

275 300 325

0

0

10.0

8.0

6.0

4.0

2.0

1.0

.8

.6

.4

.2

.1

.08

.06

.04

I I

- (r= 1.6)

6m

(r= 1.6) 6m (r 1.6)

2.1m

(r= 2.13)

(r= 2.13) 0.21 m

0.6m .. 2

m

NaIPO 4 Ratio = r

Phosphate Concentration

in Molality = m .02

.01 I I I I I I • B . . . .

525 550 575 600 625 650

Temperature (OF)

38712-170.7M

FIGURE A.2. Temperature Dependence of the Corrosion Rate of

Alloy 600 Tubing Exposed to Deaerated Phosphate

Solutions (Angwin et al. 1985).

Secondary water chemistry was changed from phosphate to AVT after approxi- mately two years of operation. Table A.2 gives the number of tubes plugged for wastage degradation from 1974 through 1977 (no tubes were plugged for

wastage after 1977).

Table A.2. Number of Tubes Plugged for Wastage Degradation

Year

1974

1975

1976

1977

No. of Tubes

14

34

19

7

A.3
The data in Table A.2 suggests that time-to-initiation for phosphate wastage

was very short and growth was not immediately arrested after conversion to

AVT. As a first approximation, wastage growth rates were estimated by

dividing the numbers in Table A.1 by four years. Performing this calculation

yields the growth rates presented in Table A.3.

As expected, cold leg wastage grows more slowly than hot leg wastage.

Removal of tubes from service by plugging did not reduce the growth rate

below the rate observed in never-plugged tubes (limited data, however).

Further, the results correlate well with the other published data.

Table A.3. Estimated Growth*Rates for Surry Generator

Wastage Degradation (mils/yr)

Leg Min Ave Max

CL** .5 1.3 3

HL 1.5 3.8 6.5

HL* 1.5 2.8 4

HL** 2.5 4.4 6.5

*Tubes never plugged during service

**Tubes plugged during service

3.0 PITTING

Limited laboratory data has been developed by Combustion Engineering under a

variety of conditions (Hall 1984). Figure A.3 gives a plot of pit depth

versus exposure time for various test conditions. Table A.4 lists pit growth

rates in mils/yr calculated from the data in Figure A.3. The average pit

growth rate for all conditions was about 7.7 mils/yr.

Table A.5 shows maximum wall-loss measured from pitted (no wastage) and pitted/wastage degraded tubes from the Surry generator.

Since the in,-service inspection and tube plugging history for the Surry

generator did not identify tubes degraded by pitting, it was difficult to

determine the pit growth rate. NDE detection of pitting is masked by copper

deposits, so determination of pit growth from field NDE data is not reliable.

One possible assumption might be one year for pit initiation followed by five

years of growth. This assumption yields the results shown in Table A.6.

These results compare reasonably well with the laboratory results given

above. On the other hand, much of the recent laboratory data-suggests that

pit growth rates are highly nonlinear. Model boiler test results reported by

Hall, et al. (1987), indicated that pit growth rates were initially rapid but

were followed by stifling of further growth due to the formation of corrosion

products that covered the outer surface of the pit.

A.4
10

o3 AVT

* Fresh Water Fault

- 0 Condensate Polisher Effluent

A Sea Water Fault -A

-:21

0-

8

6

A 00

A

4_[ A 03

A

2

'A

AA*

0

0

01

0

I

m I I m B I

20 40 60 80 100 200

Exposure, Days

FIGURE A.3. Pit Depth Versus Exposure Time in Alloy 600 Tubes

Pot Boiler Tests (Hall 1984)

400

38712-170.4M

in Model and
Table A.4. Laboratory Pit Growth Rates (mils/yr)

Environ.

Fresh

Sea

CP

AVT(1 pt)

Min

1.5

2.4

1.1

4.0

Ave

8.0

8.9

4.0

4.0

Max

25.3

26.1

6.8

4.0

Table A.5. Maximum Percent Wall-Loss for Hot Leg Pitting

and Pitting/Wastage Degradati"oh of Surry Tubes

Mech-.

p**

P/W*

P/W**

P/W

Min 14

15

34

15

Ave

42

53

48

52

Max

68.

87

67

87

*Tubes never plugged during service

**Tubes plugged during service

Table A.6

Mech.

P**

P/W* P/W**

P/W

Estimated Pitting and Pitting/Wastage

Growth Rates from Surry Tubes

Min

1.4

1.5

3.4

1.5

Ave

4.2

5.3

4.8

5.2

Max

6.8

8.7

6.7

8.7

*Tubes never plugged during service **Tubes plugged during service

4.0 DENTING

The phenomenon of denting has been arbitrarily divided into two phases; the

time to fill the tube/TSP crevice gap with magnetite (dent initiation), and

the rate of dent progression. Correlations have been developed by two

investigators (Sawsochka et al. 1986; Pathania and Balakrishnan 1986) that

relate the bulk chloride concentration measured in the blowdown water, with

time to initiate a dent. These are given as equations (1) and (2) below:

Td = 41.1 [Cl]'" 9 2 8

where Td is the time to dent initiation in years and [Cl] is the chloride

concentration in ppb, and

Y' 1.99 [Cl]. 2 9 6

(1)

(2)

A.6
APPENDIX A

COMPENDIUM OF DEFECT GROWTH RATES
APPENDIX A

COMPENDIUM OF DEFECT GROWTH RATES

1.0 INTRODUCTION

The information in this appendix summarizes much of the currently available

data on the growth rate of the common defects found in steam generator

tubing. This information was compiled to assist the determination of appropriate degradation allowances for evaluation of degraded steam generator

tubing. Most of the data was obtained from recently published research

reports and for the most part represents the results of laboratory investigations conducted under severe test conditions (e.g., high applied stresses and

aggressive environments). In some cases, field data was available for

comparison with the laboratory produced data.

2.0 THINNING (WASTAGE)

Very little data on the growth rate of wastage is available. With no plants

in the U.S. presently using phosphate secondary water chemistry, this problem

is .nt a major concern. Most recent occurrences of wastage have been observed at the periphery of the first or second tube support plate on the cold

leg side of the generator. Limited data on the progression of cold leg

thinning was reported by Baum et al. (1987). These data were obtained by

monitoring the change in the eddy-current measurements on a given tube during

successive inspections and dividing by the elapsed time. Corrosion rates

were based on data from tubes with indications having reported depths 20% or

greater in two successive inspections. The calculations were based on a

total of 489 tubes in 18 steam generators. An overall degradation rate of 3

mils/y was estimated from these data.

Other published data (Angwin et al. 1985) plotted in Figures A.1 and A.2 show

that the corrosion rate depends upon the Na/P04 molar ratio and temperature,

but maximum corrosion rates never exceeded about 5.5 mils/y.

Estimates were also obtained from evaluation of Surry metallographic data

presented by Bradley et al. (1988). Maximum wall penetration of pure wastage

was measured on tubes removed from both the hot and cold leg sides of the

Surry generator. Table A.1 gives the minimum, average, and maximum percent

wall-loss measured.

The data in Table A.1 can be used to estimate growth rates by dividing by the

duration the defect mechanism is operative. The Surry generator was in

service for slightly less than six years, from March 1973 to January 1979.

A.1
2

0

CD0V

5

4

3

2

1

1.0 1.35 1.6 2.0 2.3 2.7 3.0

Na/P0 4 38712-170.1 M

FIGURE A.1. Average Corrosion Penetrations of

Phosphate Solutions at 615OF as a

Ratio (Angwin et al. 1985).

Alloy 600 in Deaerated Sodium

Function of the Na/P04 Molar

Table A.1.

Leq

CL**

HL

HL*

HL**

Maximum Percent Wall-Loss

Degraded Surry Tubes

Min

41

12

1220

Ave

11

30

22

35

for Wastage

Max

24

52

32

52

*Tubes never plugged during service

**Tubes plugged during service

A.2
(CC)

275 300 325

10.0

8.0

6.0

4.0

- I

0

0)

2.0

1.0

.8

.6

.4

.2

.1

.08

.06

.04

.02

.01

(r= 1.6)

•-(r 1.6) 6m

6m ,

(r 1.6)

2.1m

(r= 2.13)

(r= 2.13) 0.21m

0.6m

(r= 2,13) O° "

- 1.8m.°•

S Na/PO A Ratio =r

Phosphate Concentration

in Molality = m

I I I I -L I

525 550 575 600 625 650

Temperature (OF)

38712-170.7M

FIGURE A.2. Temperature Dependence of the Corrosion Rate of

Alloy 600 Tubing Exposed to Deaerated Phosphate

Solutions (Angwin et al. 1985).

Secondary water chemistry was changed from phosphate to AVT after approximately two years of operation. Table A.2 gives the number of tubes plugged

for wastage degradation from 1974 through 1977 (no tubes were plugged for

wastage after 1977).

Table A.2. Number of Tubes Plugged for Wastage Degradation

Year

1974

1975

1976

1977

No. of Tubes

14

34

19

7

A.3
The data in Table A.2 suggests that time-to-initiation for phosphate wastage

was very short and growth was not immediately arrested after conversion to

AVT. As a first approximation, wastage growth rates were estimated by

dividing the numbers in Table A.1 by four years. Performing this calculation

yields the growth rates presented in Table A.3.

As expected, cold leg wastage grows more slowly than hot leg wastage.

Removal of tubes from service by plugging did not reduce the growth rate

below the rate observed in never-plugged tubes (limited data, however).

Further, the results correlate well with the other published data.

Table A.3. Estimated Growth Rates for Surry Generator

Wastage Degradation (mils/yr)

LeS Min Ave Max

CL** .5 1.3 3

HL 1.5 3.8 6.5

HL* 1.5 2.8 4

HL** 2.5 4.4 6.5

*Tubes never plugged during service

**Tubes plugged during service

3.0 PITTING

Limited laboratory data has been developed by Combustion Engineering under a

variety of conditions (Hall 1984). Figure A.3 gives a plot of pit depth

versus exposure time for various test conditions. Table A.4 lists pit growth

rates in mils/yr calculated from the data in Figure A.3. The average pit

growth rate for all conditions was about 7.7 mils/yr.

Table A.5 shows maximum wall-loss measured from pitted (no wastage) and pitted/wastage degraded tubes from the Surry generator.

Since the in-service inspection and tube plugging history for the Surry

generator did not identify tubes degraded by pitting, it was difficult to

determine the pit growth rate. NDE detection of pitting is masked by copper

deposits, so determination of pit growth from field NDE data is not reliable.

One possible assumption might be one year for pit initiation followed by five

years of growth. This assumption yields the results shown in Table A.6.

These results compare reasonably well with the laboratory results given

above. On the other hand, much of the recent laboratory data suggests that

pit growth rates are highly nonlinear. Model boiler test results reported by

Hall, et al. (1987), indicated that pit growth rates were initially rapid but

were followed by stifling of further growth due to the formation of corrosion

products that covered the outer surface of the pit.

A.4
10

3 AVT

0 Fresh Water Fault

- 0 Condensate Polisher Effluent

A Sea Water Fault

8

C-" 0S

A

cn

6

4

A

A

A

2

A

AA 0

0

AA

0

0o

S

S

I

20 40 60 80 100 200 400

Exposure, Days 38712-170.4M

Pit Depth Versus Exposure Time in Alloy 600 Tubes in Model and

Pot Boiler Tests (Hall 1984)

FIGURE A.3.
Table A.4. Laboratory Pit Growth Rates (mils/yr)

Environ.

Fresh

Sea

CP

AVT(1 pt)

Min

1.5

2.4

1.1

4.0

Ave

8.0

8.9

4.0

4.0

Max

25.3

26.1

6.8

4.0

Table A.5. Maximum Percent.Wall-Loss for Hot Leg Pitting

and Pitting/Wastage Degradation of Surry Tubes

Mech.

P**

P/W* P/W**

P/W

Min 14

15

34

15

Ave

42

53

48

52

Max

68

87

67

87

*Tubes never plugged during service

**Tubes plugged during service

Table A.6

Mech. p**

P/W*

P/W** P/W

Estimated Pitting and Pitting/Wastage

Growth Rates from Surry Tubes

Miný

1.4

1.5

3.4

1.5

Ave

4.2

5.3

4.8

5.2

Max

6.8

8.7

6.7

8.7

*Tubes never plugged during service **Tubes plugged during service

4.0 DENTING

The phenomenon of denting has been arbitrarily divided into two phases; the

time to fill the tube/TSP crevice gap with magnetite (dent initiation), and

the rate of dent progression. Correlations have been developed by two

investigators (Sawsochka et al. 1986; Pathania and Balakrishnan 1986) that

relate the bulk chloride concentration measured in the blowdown water, with

time to initiate a dent. These are given as equations (1) and (2) below:

Td = 41. 1[C1-'928

where Td is the time to dent initiation in years and [Cl] is the chloride

concentration in ppb, and

Y' 1.99 [Cl]" 2 9 6

(1)

(2)

A.6
where Y' is the corrosion rate in mils/y. Equation (2) can be rearranged to

give an equation of the same form as equation (1):

Td = .503 Yr [Cl]-' 2 9 6 (3)

where Yr is the crevice gap in mils.

These correlations are only applicable to plants that are sea/brackish water

cooled and with AVT secondary water chemistry. In the development of equation (1) the crevice gap was assumed to be 7.5 mils. Substituting this value

into equation (3) yields a coefficient of 3.77. For a chloride concentration

of 200 ppb (represents a crude estimate of the low end of Surry conditions),

equation (1) predicts a time-to-denting of about 4 months and equation (2)

predicts about 9 months. These values seem reasonable given the experience

at Surry (Doctor et al. 1983).

Of course, the time-to-denting depends on factors in addition to the bulk

chloride concentration. The presence of copper in the secondary water has

been observed to increase corrosion by lowering the crevice pH. Heat transfer test results clearly show the corrosion rate and hence time-to-denting

strongly depend on the crevice superheat. In addition, tube support plate

design and material of construction can change the corrosion rate. More open

geometries such trefoil, quatrefoil, 'egg crate' and lattice bars are less

susceptible to accelerated corrosion compared with the cylindrical annuli in

drilled support plates. This may be explained by the lower crevice superheat

in the more open geometries which results in lower concentrations of

contaminants in the crevice and hence lower corrosion rates. Support plates

made from Type 405 stainless steel are more corrosion resistant than carbon

steel. Data from single tube model boiler tests in sea water

indicated that the corrosion rate of Type 405 stainless steel support plates

was about 5-10 times lower than carbon steel. However, at high bulk chloride

concentrations (30 ppm) Types 405 and 409 tube supports corroded at rates

comparable to carbon steel.

The available evidence on dent progression suggests the corrosion rate

decreases once the crevice gap is filled. Reductions in corrosion rate

ranging from 3 to 10 have been observed. Thus radial dent progression for a

given chloride concentration may be predicted by multiplying the initial

corrosion rates by a factor between 0.1 to 0.3. If a chloride concentration

of 200 ppb is substituted into equation (2), then corrosion rates of about

9.4 mils/yr are predicted. This translates to dent progression rates of

about 1 to 3 mils/yr.

The Surry plant was on AVT chemistry for four years. If the crevice gap

filled within 4 to 9 months following conversion to AVT, then between 3 and

11 mils of radial denting would be predicted (assuming a constant denting

rate). This would result in tube diameter changes of between 6 and 22 mils.

Eddy-current profilometry measured an average diameter decrease of approximately 15 mils on the cold leg side and about 45 mils on the hot leg side.

These results are in fair agreement with the predicted values from the bulk

blowdown chemistry correlations. Better agreement would not be expected

since it is difficult to derive a single chloride concentration that

A.7
represents Surry operating history. It was not uncommon for the blowdown

chemistry to exhibit large periodic increases in chloride. In one transient

up to 50,000 ppb was measured (Doctor et al. 1983).

4.1 CORRELATION OF DENTING STRAIN WITH IGSCC INITIATION

Electromechanical profilometry measurements were made on 101 tubes on the hot

leg side of the Surry generator. The maximum principal strain was computed

from these data by analyzing the shape of the dented tube cross

section. A histogram of the strain measurement results is given in Figure A.4.

The average measured strain was about 7.4% for the entire generator. If the

denting at the top of tube sheet is excluded, then the average increases to

8.1%. Assuming that denting initiated within 4 to 9 months after conversion

to AVT chemistry, the average strain rate would be about 7 x 10-10 .sec-1 . Constant extension rate tests (CERT) conducted by Bandy and Van Rooyan (1984)

at a strain rate of 2 X 10-7 sec- 1 have shown decreasing plastic strain

required for SCC initiation with increasing temperature, Figure A.5. By

extrapolating Van Rooyan's data to 288%C (550°F) and 316°C (600 0 F), estimates

of about 12% and 7.5% plastic strain are obtained for operating conditions.

Thirty tubes with calculated strains ranging from 4.2% to 88.3% were removed

from the hot leg side of the Surry generator to determine the frequency of

SCC with increasing strain. These tubes were examined visually and in some

cases metallographically. Results of these examinations can be divided into

three strain regions: 0% to 10%, 10% to 20% and greater than 20%. No

cracking was observed below 10%, while approximately one-third of. the

specimens with calculated strain levels between 10% and 20% exhibited ID

cracking. All of the specimens above 20% were found to contain ID cracks.

These results compare reasonably well with the predicted values based on the

CERT data. These comparisons are only semi-quantitative, since material

susceptibility, temperature and strain rate all play an important role in

determining the propensity for SCC.

5.0 INTERGRANULAR ATTACK AND STRESS CORROSION CRACKING

Intergranular attack (IGA) and secondary water stress corrosion cracking

(SWSCC) have been observed in at least 30 domestic and foreign reactors.

SWSCC is characterized by single or multiple major cracks that affect only a

few grain boundaries. In Contrast, volumetric IGA affects all the grain

boundaries and may be accompanied by deeper SWSCC-type penetrations.

A significant number of plants have also been affected by primary water

stress corrosion cracking (PWSCC). High stress, either applied, residual or

a combination of the two appears to be a necessary condition for PWSCC.

Cyclic load and the associated strain rate may reduce the level of stress

required for cracking. Higher temperature tends to increase cracking rates.

The bulk of the laboratory data for Alloy 600 in PWR environments is related

to crack initiation. Most investigations have utilized specimens such as.

tensile bars (CERT), bent beams, U-bends or C-rings. Typically, these

specimens have been loaded or bent, exposed in a controlled environment

A.8
500-

400-

EnQ

0

a)LL_

o

C)o Lf o

300-

200-

428

V/i

262

.' .

100-

0 5

14 8 9 4 3 1 3 3

-- r I, I--- r -T--rT---r -- r- .-...

15 20 25 30 35 40 4,5 50 55 60 65 70

Percent Strain at Tube Sheet,,"Support Plates

2 1 1

75 80 85 90 95 100

FIGURE A.4. Histogram Showing Strain Measurements From the PROFIL-360 Inspection (Doctor et a]. 1988b)
20

10

9

T

(I,

1 , I I I i 280 300 320 340 360 380

Temperature, 'C

FIGURE A.5. Temperature Dependence of SCC Initiation Strain for CERT Tests

at 6 = 2 x 10-10 sec- 1 . (Reprinted with permission from

Corrosion, Vol. 40, No. 8, "Stress Corrosion Cracking of

Inconel Alloy 600 in High Temperature Water - An Update."

Bandy and Van Rooyan, National Association of Corrosion Engineers,

Houston, Texas.)

A. 10
and periodically removed from the environment for examination. The results

of these tests are usually the time-to-crack or time-to-failure. In some

cases, estimates of the crack growth rate were obtained. Very few studies

employed fracture mechanics type specimens, with sharp starter notches for

measuring crack growth. Usually temperature, stress, or strain rate are used

as accelerating parameters. Often, the test conditions were severe compared

to expected service conditions.

The growth rates of IGA and SCC can only be estimated approximately for the

reasons noted above. Figures A.6 and A.7 summarize the available data (Bandy

and Van Rooyan 1984; Partridge 1987; Hermer and Wolf 1986; Aspden 1986;

Newman et al. 1987; Theus 1981; Jacko 1986) on growth of IGA and SCC (both primary and secondary) as a function of temperature. Note the strong

temperature dependence of all three forms of cracking. Another interesting observation is that SCC is more rapid in primary than in secondary water.

Although not shown in Figures A.6 and A.7, both IGA and SWSCC appear to be

insensitive to caustic concentration over the range 10 to 40%. Below 10%

caustic, the growth rate decreases with decreasing concentration. Two

additional observations not explicitly depicted in Figure A.7 are that

different heats of material behave differently and little difference is

observed in the PWSCC growth rate in pure water, primary water, or AVT

treated water. Also shown in Figure A.7 is the decreased rate of SWSCC for

thermally treated (TT) Alloy 600 relative to mill annealed (MA) material.

Corrosion resistance to IGA is improved by thermal treatment by a factor of

about 1.6.

Figures A.8 and A.9 show the limited amount of data (Newman et al. 1987 and

Berge 1986) suggesting a relationship between applied stress (or stress

intensity) and growth rate for IGA and SWSCC respectively. For SWSCC at high applied stress levels, the curves in Figure A.8 suggest a saturation of crack

growth rate. Also a significant dependence on caustic concentration was

observed over the full range of stress intensity factors investigated. In

addition, the results in Figure A.8 suggest a threshold below which no

significant SWSCC is expected.

Upper bound estimates for growth rates at 550°F and 600°F obtained from

Figures A.6 and A.7 are listed in Table A.7.

A.11
1

0.1

00

0 -o

E V0,I

Cr

0.01o ED

< • •o O Capsules, 10% NaOH (Partridge 1.987)

-Capsules, 25% NaOH (Partridge 1987)

j C-Rings,..9% NaOH/1% KOH

0 I(Partridge 

1 987)

v .C-Rings, 10 & 50% Caustic (Partridge 1987)

O * Capsules, 10 & 50% Caustic (Partridge 1987)

0.001 MC-Ring 40% NaOH/10% KOH/Cr 2 0 3

V •(Hermer, et al. 1986)

0 Capsules, 40/10, 20/5, 8/2, 4%, 1%

NaOH/KOH (Aspden 1986)

VModel Boiler (Partridge 1987)

-C-Rings 10-40% NaOH (Newman,et al. 1987)

-- ,-Capsules, Calculation (Partridge 1987)

0.0001 , I I I

280 300 320 340 360 380

Temperature, 'C

FIGURE A.6. Temperature Dependence of IGA Growth Rate

A.12
10 U

A

0

0

0

0

E0

A

a

0 1 kCO

0

0~

UU

C')

0D

0.1

)

B

/

03

/

/

/

/

0.01

0.001-

280

O MA, Pure or PWR Water

(Bandy et al. 1984)-

O Capsules, 10% NaOH

(Partridge 1987)

<>Capsules, 25% NaOH +

Carbonate (Partridge 1987)

*Capsules 25% NaOH + Sulfate

(Partridge 1987)

MA C-Ring, 10% NaOH

(Jacko 1986)

TT C-Ring, 10% NaOH

(Jacko 1986)

A MA RUB, Pure Water

(Theus 1981)

\_0 \eC

300 320 340 360 380

Temperature, 'C

FIGURE A.7. Temperature Dependence of SCC Growth Rate

A. 13
15

10

U)

05

0

5

0

0 100 200 300 400 500

38712-170.6M

Stress, MPa

FIGURE A.8. Influence of Stress on IGA as a Function

of Temperature (Newman et al. 1987).

Table A.7. Estimated Upper Bound Growth Rates

for Cracking Mechanisms (mils/yr)

Mechanism

IGA

SWSCC

PWSCC

5500F

5

50

170

6000 F

30

85

600

A.14
daT- (Um/h) dt

10

1

10

1

0.1

Alloy 69(

Alloy 69(

4

MA - NaOH 100 Q/1

'L

HT - NaOH 100 g/l

0.1 4

4

4 Alloy 690 HT - NaOH 4 gq1

0 10 20 30 40 50 60 K1 (MPa 47-) 0 10 20 30 40 50 60 K1 (MPa"T)

MA: Mill Annealed

HT: Heat Treated 16h at 700 0C

38712-170.5M

FIGURE A.9. Stress Corrosion Tests in Deaerated Sodium Hydroxide at 350'C on Fracture Mechanics

Type Specimens: Comparison of Alloy 600 and 690 Behavior. Effect of Heat Treatment

at 700'C. (Berge 1986)
6.0 CORROSION FATIGUE

For corrosion fatigue conditions the material property needed is the cyclic

crack growth rate, da/dN. The data on Alloy 600 in typical PWR environments

is limited. Furthermore, for the low K regime, there are few da/dN data in

any environment. Figure A.1O (Davis et al. 1983) shows a summary of the

crack growth rate results for Alloy 600 in air and water environments. These

data combined with appropriate stress analyses can be used to estimate crack

growth for tubes in OTSGs affected by this damage mechanism.

7.0 SUMMARY

The foregoing material briefly presented much of the publicly available

information on growth rates of various damage mechanisms for steam generator

tubing. The information is sparse in most instances and nonexistent in other

cases. Table A.8 provides upper bound growth rate estimates at 600OF for the

major damage mechanisms of concern.

Table A.8. Upper Bound Growth Rates (600*F) for Principal

Steam Generator Tube Damage Mechanisms (mils/yr)

Mechanism Rate

Thinning 6

Pitting 25

IGA 30

SWSCC 85

PWSCC 600

It is clear the damage mechanism most likely to penetrate the tube wall

between inspection outages is primary side cracking. However, all damage forms (except for thinning) potentially grow at rates that could penetrate

the tube wall during a typical eighteen month period between inservice

inspections. This assumes that a particular damage mechanism operates at a

constant rate with time, which as Figure A.11 illustrates may not be the case

for IGA (Miglin 1987). It is typical of non-stress-assisted corrosion mechanisms to exhibit decreasing rate with time. However, not enough data exists

to permit more accurate defect growth rate estimates for many of the damage

mechanisms discussed here.

A.16
10,

5 5 4 / I

'8/ 6/

I!1i

S/ III

7// - I2

101 10 -7 -

5

4 5 6 8 10 20 30 40 50

Stress Intensity Factor Range, A K (ksi -in.)

Key to crack growth test conditions:

1 400 cpm in air at 300 OF, R = .05

2 400 cpm in air at 800 0 F, R = .333

3 400 cpm in air at 300 0 F, R =.6

4 40 cpm in air at 75 OF, R = .05

5 40 cpm in air at 1200 OF, R = .05

6 900 cpm in pure water at 185 OF, 15 ppm 0 2, R = 0, solution annealed in air

7 1 cpm in pure water at 185 0 F, 15 ppm 02 , R = 0, solution annealed

8 1 cpm in pure water at 185 0 F, 15 ppm 02 ,R = 0, sensitized 24h at 650 0C

9 1 cpm in pure water at 185 0 F, 15 pprn 02 , R = 0, sensitized 1000h at 650 OC

38712-170.2

FIGURE A.10. Summary of Corrosion Fatigue Crack Growth

Results for Inconel 600 (Davis 1983)

A.17
10.00

8.00

6.00

4.00

M

E

0

._

0-

0_

3 (Time)0 .

6 3

.8 (Time) 0 "

6 0

38712-170.3M

2.00

0.00

0 2000 4000 6000 8000

Exposure Time, hr

FIGURE A.11. Depth of IGA Penetrations on Mill-Annealed Alloy 600

Tubing (Heat NX3335) as a Function of Exposure Time

(Miglin 1987)

8.0 REFERENCES

Angwin, M. J. et al. 1985.

Research Institute, Palo

Steam Generator Reference Book. Electric Power

Alto, California.

Aspden, R.

Devices

Corrosi

Tubing.

Califor

G. 1986. "Background Reporting and IGA Screening in Static

In Proceedings: 1984 Workshop on Secondary Side Stress

on Cracking and Intergranular Corrosion of PWR Steam Generator

EPRI NP-4478, Electric Power Research Institute, Palo Alto,

nia.

Test

Bandy, R. and D. Van Rooyan. 1984. "Stress Corrosion Cracking

Alloy 600 in High Temperature Water - An Update." Corrosion

40(8) :425-430.

of Inconel

Baum, A. J., C. Dangelo, P. J. Kuchirka, N. Singleton, and M. J. Sredzienski.

1987. Steam Generator Cold Leg Thinning in Operating Plants. EPRI

NP-5140, Electric Power Research Institute, Palo Alto, California.

Berge, J. P. 1986. "Questions on U-Bend Cracking." In Proceedings: 1984

Workshop on Secondary Side Stress Corrosion Cracking and Intergranular

Corrosion of PWR Steam Generator Tubing, EPRI NP-4478, Electric Power

Research Institute, Palo Alto, California.

A.18
Bradley, E. R., P. G. Doctor, R. H. Ferris, and J. A. Buchanan. 1988. Steam

Generator Group Project Task 13 Final Report - NDE Validation.

NUREG/CR-5185, PNL-6399, Pacific Northwest Laboratory, Richland,

Washington.

Davis, C. S., J. M. Thomas, S. W. Winder, and D. E. Allison. 1983. An

Engineering and Probabilistic Analysis of Tube Cracking Performance in

Once-Through Steam Generators. EPRI NP-3065, Electric Power Research

Institute, Palo Alto, California.

Doctor, P. G., J. M. McIntyre, P.K. Alley, L. R. DesCamp, J. S. Littlefield,

C. A. LoPresti, and J. A. Buchanan. 1983. The In-Service History of

The Surry Unit 2A Steam Generator. PNL-4880, Pacific Northwest

Laboratory, Richland, Washington.

Doctor, P. G., H. Harty, R. H. Ferris, and A. S. Birks. 1989. Steam

Generator Group Project Task 9 Final Report - Nondestructive Evaluation

Round, Vol. 1: Description and Summary Data. NUREG/CR-4849, PNL-5868,

Pacific Northwest Laboratory, Richland, Washington.

Hall, J. F. 1984. "Pitting in the Laboratory at Combustion Engineering."

In Workshop Proceedings: Pitting in Steam Generator Tubing. EPRI

NP-3574-SR, Electric Power Research Institute, Palo Alto, California.

Hall, J. F., R. S. Frisk, K. E. Marugy, and A. S. O'Neill. 1987.

Investigation of Causes and Corrective Actions for Pitting in Steam

Generator Tubes: Prototypic Tests. EPRI NP-5248, Electric Power Research

Institute, Palo Alto, California.

Hermer, R. E. and C. R. Wolfe. 1986. "Alloy 600 Intergranular

Corrosion in Alkaline Environments." In Proceedings: 1984 Workshop

on Secondary Side Stress Corrosion Cracking and Intergranular

Corrosion of PWR Steam Generator Tubing. EPRI NP-4478, Electric Power

Research Institute, Palo Alto, California.

Jacko, R. J. 1986. "IGSCC and IGA Behavior of Inconel 600." In

Proceedings: 1984 Workshop on Secondary Side Stress Corrosion

Cracking and Intergranular Corrosion of PWR Steam Generator Tubing.

EPRI NP-4478, Electric Power Research Institute, Palo Alto, California.

Miglin, B. P. 1987. Production of Eddy-Current Standards for Caustic

Intergranular Corrosion. EPRI NP-5109, Electric Power Research Institute,

Palo Alto, California.

Newman, J. R., P. McIntyre, and S. M. Payne. 1987. "Effect of Thermal

Treatment and Impurity Segregation on the Kinetics of IGA in Inconel

600." Presented at EPRI Workshop on IGA and SCC Mechanisms, Electric

Power Research Institute, Palo Alto, California.

Partridge, M. J. 1987. "Effect of Temperature and Caustic on IGA and

IGSCC Corrosion Rates." Presented at EPRI Workshop on IGA and SCC

Mechanisms, Electric Power Research Institute, Palo Alto, California.

A.19
Pathania, R. S. and P. V. Balakrishnan. 1986. Correlation of Tube

Support Structure Studies. EPRI NP-4672, Electric Power Research

Institute, Palo Alto, California.

Sawsochka, S. G., K. D. Friedman, S. S. Choi, and J. K. Pyo. 1986.

Correlation of Tube Support Corrosion Studies. EPRI NP-4818, Electric

Power Research Institute, Palo Alto, California.

Theus, G. J. ,1981. "Stress Corrosion Cracking Tests of Alloy 600. In

Workshop Proceedings: U-Bend Cracking in Steam Generators,

EPRI WS-80-136, Electric Power Research Institute, Palo Alto, California.

A.20
APPENDIX B

STEAM GENERATOR TUBE INTEGRITY PROGRAM/STEAM GENERATOR GROUP PROJECT

REPORTS AND PUBLICATIONS
APPENDIX B

STEAM GENERATOR TUBE INTEGRITY PROGRAM/STEAM GENERATOR GROUP PROJECT

REPORTS AND PUBLICATIONS

Alzheimer, J. M., R. A. Clark, W. E. Anderson, J. C. Crowe, and M. Vagins.

1977. "Steam Generator Tube Integrity Program. Quarterly Report, January

1-March" 31, 1977." NUREG-0359. Pacific Northwest Laboratory, Richland,

Washington.

Alzheimer, J. M., R. A. Clark, C. J. Morris and M. Vagins. 1979. "Steam

Generator Tube Integrity Program Phase I Report." NUREG/CR-0718,

PNL-2937, Pacific Northwest Laboratory, Richland, Washington.

Morris, C. J., G,. H. Lyon, T. J. Davis and C. B. Perry. 1980. "Eddy-Current

Inspection of INCONEL-600 Steam Generator Tubes at the Tube Sheet."

NUREG/CR-1626, PNL-3472, Pacific Northwest Laboratory, Richland,

Washington.

Clark, R. A. and V. PitzPatrick. 1981. "Steam Generator Transport: Surry,

Virginia to Hanford, Washington," p. 826-827 in Trans. Am. Nucl. Soc.

(United States), Vol. 39.

Wheeler, K. R. and R. A. Clark. 1982. "Initial Inspection of a Service

Degraded Steam Generator Removed from Service," p. 297-298 in Trans. Am.

Nuc. Soc. (United States), Vol. 41.

Clark, R. A. and M. Lewis. 1983. "Repair Research as Part of the Steam

Generator Group Project," p. 493 in Trans. Am. Nucl. Soc. (United States),

Vol. 44.

Clark, R. A. and M. Lewis. 1983. "Steam Generator Group Project Semi-Annual

Progress Report, January - June 1983." PNL-4859, Pacific Northwest

Laboratory, Richland, Washington.

Doctor, P. G., J. M. McIntyre, P. K. Alley, L. R. DesCamp, J. S. Littlefield,

C. A. LoPresti and J. A. Buchanan. 1983. "The In-Service History of the

Surry Unit 2A Steam Generator." PNL-4880, Pacific Northwest

Laboratory, Richland, Washington.

Reece, W. D., G. R. Hoenes and M. A. Parkhurst. 1984. "Steam Generator Group

.Project Progress Report - Task 3 Health Physics." NUREG/CR-3578,

PNL-471'1, Pacific Northwest Laboratory, Richland, Washington.

Doctor, P. G., J. A. Buchanan, J. M. McIntyre, P. J. Hof and

S. S. Ercanbrack. 1984. "Steam Generator Group-;Project Progress Report

on Data Acquisition/Statistical Analysis." NUREG/CR-3579, PNL-3955,

Pacific Northwest Laboratory, Richland, Washington.

B.1
Allen, R. P., R. L. Clark and W. D. Reece. 1984. "Steam Generator Group

Project Task 6 - Channel Head Decontamination." NUREG/CR-3841, PNL-4712,

Pacific Northwest Laboratory, Richland, Washington.

Clark, R. A. and M. Lewis. 1984. "Steam Generator Group Project Semi-Annual

Progress Report July - December 1982." NUREG/CR-3580, PNL-4692, Pacific

Northwest Laboratory, Richland, Washington.

Clark, R. A. and M. Lewis. 1984. "Steam Generator Group Project Annual

Report - 1982." NUREG/CR-3581, PNL-4693, Pacific Northwest Laboratory,

Richland, Washington.

Bickford, R. L., R. A. Clark, P. G. Doctor, H. E. Kjarma and C. A. LoPresti.

1984. "Eddy Current Round Robin Test on Laboratory Produced Intergranular

Stress Corrosion Cracked Inconel Steam Generator Tubes." NUREG/CR-3561,

PNL-4695, Pacific Northwest Laboratory, Richland, Washington.

Clark, R. A. and R. L. Bickford. 1984. "Steam Generator Tube Integrity

Program Leak Rate Tests - Progress Report." NUREG/CR-3562, PNL-4629,

Pacific Northwest Laboratory, Richland, Washington.

Wheeler., K. R., P. G. Doctor, L. K. Fetrow and M. Lewis. 1984. "Steam

Generator Group Project Task 8 - Selective Tube Unplugging."

NUREG/CR-3842, PNL-4876, Pacific Northwest Laboratory, Richland,

Washington.

Schwenk, E. B. and K. R. Wheeler. 1984,. "Steam Generator Group Project

Task 10 - Secondary Side Examination." NUREG/CR-3843, PNL-5033, Pacific

Northwest Laboratory, Richland, Washington.

Sinclair, R. B. 1984. "Secondary Side Photographic Techniques Used in

Characterization of Surry Steam Generator." NUREG/CR-3094, PNL-5053,

Pacific Northwest Laboratory, Richland, Washington.

Tanner, J. E., W. D. Reece and R. I. Scherpelz. 1984. "Steam Generator Group

Project Progress Report - Task 3." PNL-5269, Pacific Northwest

Laboratory, Richland, Washington.

Wheeler, K. R. 1984. "Initial Inspection of the Service Degraded Surry 2A

Steam Generator, June 1982". PNL-5268, Pacific Northwest Laboratory,

Richland, Washington.

Clark, R. A., M. Lewis and J. Muscara. 1984. "Steam Generator Group Project

Semi-Annual Progress Report January Through June 1984." PNL-5267, Pacific

Northwest Laboratory, Richland, Washington.

Birks, A. S., R. H. Ferris, P. G. Doctor and H. Harty. 1984. "Preliminary

Report on Steam Generator Group Project - Task 9 NDE Round Robin."

PNL-5499, Pacific Northwest Laboratory, Richland, Washington.

B.2
Clark, R. A. et al. 1984. "1983 Branch Annual Report - Steam Generator

Integrity Program." NUREG-0975, Vol. 2, PNL-SA-11720, Pacific Northwest

Laboratory, Richland, Washington.

Clark, R. A. 1984. "Status and Progress of Research on a Removed-from-Service

Steam Generator." NUREG/CP-0048, Vol. 4, p. 98-116, in Proceedings of

the Eleventh Water Reactor Safety Research Information Meeting, Pacific

Northwest Laboratory, Richland, Washington.

Clark R. A. 1984. "Results of Research Utilizing a Retired from Service PWR

Steam Generator." PNL-SA-12531, presented at Specialist Meeting on Steam

Generator Problems, Stockholm, Sweden, October 1, 1984.

Clark, R. A. 1985. "Status and Progress of Research on a Removed-from-Service

Steam Generator," p. 39-47, Nuclear Engineering and Design. Pacific

Northwest Laboratory, Richland, Washington.

Clark, R. A. 1985. "Characterization of the Removed-from-Service Surry 2A

Steam Generator." PNL-SA-12532, presented at 8th International

Conference on Structural Mechanics in Reactor Technology, Brussels,

Belgium, August 19, 1985.

Clark, R. A. 1985. "Metallurgical Analysis and Comparative Nondestructive

Inspection Results on Specimens Removed from a Retired from Service PWR

Steam Generator," International Symposium on Contribution of Materials

Investigation to the Resolution of Problems Encountered in PWR Plants,

Chinon, France, September 3, 1985.

Clark, R. A. and M. Lewis. 1985. "Observation of Denting and Other

Deterioration in the Surry Steam Generator," PNL Report PNL-SA-13273,

presented at American Society for Metals Meeting, San Diego, California,

July 9, 1985.

Clark, R. A. and M. Lewis. 1985. "Observation of Denting and Other

Deterioration in the Surry Steam Generator," p. 193-196 in Proceedings

of the International Conference on Nuclear Power Plant Aging,

Availability Factor and Reliability Analysis (American Society of Metals,

Metals Park, Ohio).

Clark, R. A., E. R. Bradley, and R. J. Kurtz. 1985. "Investigation of Tube

Sheet Samples Removed from the Retired Surry Steam Generator." PNL-SA13467, presented at Workshop on Remedial Action for Intergranular

Corrosion, Baltimore, Maryland, October 15, 1985.

Clark, R. A. and P. G. Doctor. 1985. "Surry Steam Generator Program and

Baseline Eddy Current Examination," Vol. 4, p. 260-285, in Proceedings

of the Twelfth Water Reactor Safety Research Information Meeting.

NUREG/CP-0058, Vol. 4. Pacific Northwest Laboratory, Richland,

Washington.

B.3
Clark, R. A. and M. Lewis. 1985. "Steam Generator Group Project Annual

Report - 1983." NUREG/CR-4361, PNL-5017, Pacific Northwest Laboratory,

Richland, Washington.

Clark, R. A. and M. Lewis. 1985. "Steam Generator Group Project Annual

Report - 1984." NUREG/CR-4362, PNL-5417, Pacific Northwest Laboratory,

Richland, Washington.

Clark, R. A. and M. Lewis. 1985. "Steam Generator Group Project Semi-Annual

Progress Report, January through June 1985." PNL-5603, Pacific Northwest

Laboratory, Richland, Washington.

Clark, R. A., R. L. Bickford, A. S. Birks, R. L. Clark, P. G. Doctor, R. H.

Ferris, L. K. Fetrow, W. D. Reece, E. B. Schwenk, and G. E. Spanner.

1985. "1984 Branch Annual Report - Steam Generator Integrity Program."

NUREG-0975, Vol. 3, PNL-SA-13015, Pacific Northwest Laboratory, Richland,

Washington.

Clark, R. A., P. G. Doctor, and R. H. Ferris. 1986. "Surry Steam Generator -

Examination and Evaluation," Vol. 2, p. 409-430 in Proceedings of the

Thirteenth Water Reactor Safety Research Information Meeting, NUREG/CP0072, Vol. 2. Pacific Northwest Laboratory, Richland, Washington.

Kurtz, R. J. and M. Lewis. 1986. "Steam Generator Group Project Semi-Annual

Progress Report, January through June 1986." PNL-6012, Pacific Northwest

Laboratory, Richland, Washington.

Kurtz, R. J. and R. A. Clark. 1986. "1985 Branch Annual Report - Steam

Generator Integrity Program." NUREG-0975, Vol. 4, PNL-5772, Pacific

Northwest Laboratory, Richland, Washington.

Clark, R. A., P. G. Doctor, and R. H. Ferris. 1987. "Surry Steam Generator -

Examination and Evaluation," p. 123-134, Nuclear Engineering and Design.

Kurtz, R. J., M. Lewis and R..A. Clark. 1987. "Steam Generator Group Project

Annual Report - 1985." NUREG/CR-4848, PNL-5771, Pacific Northwest

Laboratory, Richland, Washington.

Kurtz, R. J. 1987. "Using a Retired Surry Unit to Assess ISI (In-Service

Inspection) Reliability," Nuclear Engineering and Design, p. 37-39.

Schwenk, E. B. 1987. "Steam Generator Group Project Task 10 - Secondary

Side Examination Final Report." NUREG/CR-4850, PNL-6045, Pacific

Northwest Laboratory, Richland, Washington.

Kurtz, R. J. and M. Lewis. 1987. "Steam Generator Group Project Annual

Report - 1986." PNL-6150, Pacific Northwest Laboratory, Richland,

Washington.

Kurtz, R. J. 1987. "1986 Branch Annual Report - Steam Generator Integrity

Program." NUREG-0975, Vol. 5, PNL-6110, Pacific Northwest Laboratory,

Richland, Washington.

B.4
Bradley, E. R., P. G. Doctor, R. H. Ferris, and J. A. Buchanan. 1988.

Generator Group Project - Task 13 Final Report - NDE Validation."

NUREG/CR-5185, PNL-6399, Pacific Northwest Laboratory, Richland,

Washington.

"Steam

Doctor, P. G., H. Harty, R. H. Ferris and A. S. Birks. 1988. "Steam

Generator Group Project - Task 9: Final Report - Nondestructive Evaluation Round Robin, Comments Vol. I: Description and Summary Data and Vol.

II." NUREG/CR-4849 Vol. I and II, PNL-5868, Pacific Northwest Laboratory,

Richland, Washington.

'Doctor, P. G., A. S. Birks, R. H. Ferris, H. Harty and G. E. Spanner.

1988. "Steam Generator Group Project - Task 7: Final Report - PostService Baseline Eddy-Current Examination." NUREG/CR-5087, PNL-5940,

Pacific Northwest Laboratory, Richland, Washington.

Kurtz, R. J., and R. A. Clark. 1988. "Compendium and Comparison of

International Practice for Plugging, Repair, and Inspection of Steam

Generator Tubing." NUREG/CR-5016, CSNI No. 140, PNL-6341, Pacific

Northwest Laboratory, Richland, Washington.

Kurtz, R. J., J. M. Alzheimer, R. L. Bickford, R. A. Clark, C.

E. A. Simonen and K. R. Wheeler. 1988. "Steam Generator

Program Phase II Final Report." NUREG/CR-2336, PNL-4008,

Northwest Laboratory, Richland, Washington.

J. Morris,

Tube Integrity

Pacific

Kurtz, R. J. et al. 1988. "1987 Branch Annual Report - Steam Generator

Integrity Program." NUREG-0975, Vol. 6, PNL-6401, Pacific Northwest

Laboratory, Richland, Washington.

Kurtz, R. J., R. L. Bickford, W. M. Bowen, E. R. Bradley, P. G. Doctor, R. H.

Ferris, L. K. Fetrow, and F. A. Simonen. 1988. "Surry Steam Generator

Program - Final Results - Evaluation of NDE Performance and Technical

Basis for Improved ISI and Plugging Criteria," Vol. 2, p. 429-457, in

Proceedings of the Fifteenth Water Reactor Safety Research Information

Meeting, USNRC Conference Proceeding NUREG/CP-0091, Vol. 2. Pacific

Northwest Laboratory, Richland, Washington.

Bowen, W. M., P. G. Heasler, and R. B. White. 1989. "Evaluation of

Sampling Plans for In-Service Inspection of Steam Generator Tubes:

Part I." NUREG/CR-5161, PNL-6462, Pacific Northwest Laboratory, Richland,

Washington.

Kurtz, R. J., R. A. Clark, E. R. Bradley, W. M. Bowen, P. G. Doctor,

F. A. Simonen, and R. H. Ferris. 1990. "Steam Generator Tube Integrity

Program/Steam Generator Group Project - Final Summary Report."

NUREG/CR-5117, PNL-6446, Pacific Northwest Laboratory, Richland,

Washington.

B.5
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NRC FORM 335 U.S. NUCLEAR RECULATORY COMMISSION 1. REPORT NUMBER F., 89 (Assigned by NRC, Add Vol., Supp., Rev., NR(ý- 10", and Addendum Numbers, if era .) 320! 3202 BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse) NUREG/CR-5117

2. TITLE AND SUBTITLE PNL-6446

Steam Generator Tube Intecirity Prooram/ 3 DATE REPORT PUBLISHED

Steam Generator Group Project MONTH R EAR

Final Project Summary Report May 1990

4. FIN OR GRANT NUMBER

FIN B2097 5. AUTHOR(S) 6. TYPE OF REPORT

R. J. Kurtz, R. A. Clark, E. R. Bradley, W. M. Bowen

P. G. Doctor, R. H. Ferris, F. A. Simonen Technical

7. PERIOD COVERED (lncl,.,i, -

1976 - 1988

8. PERFORMING ORGANIZATION - NAME AND ADDRESS 'if NRC, provide Division, Office or Region, U.S. Nuchlar Regulatory Con.ission andd ai,'ingaddresj, ii cont-actor, orcvid"

n-,ntt* a-d mailing address.)

Pacific Northwest Laboratory

PO Box 999

Richland, WA 99352

9. SPONSOR ING ORGANI ZATI ON - NAM E AND ADD R ESS (If NRC, type "'Sante as above " if contrractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Cormtniss ion,

and mailing address. )

Division of Enqineerinq

Office of Nuclear Regulatory Research

U.S. Nuclear Requlatory Commission

Washinqton, DC 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less)

The Steam Generator Tube Integrity Program/Steam Generator Group Project was a threephase program conducted for the U.S. Nuclear Regulatory Commission (NRC) by Pacific

Northwest Laboratory. The main goal of the program was to provide the NRC with validated information on the reliability of nondestructive examination techniques to detect

and size flaws in steam generator tubing and to determine the remaining integrity of

service-degraded tubing. The program was performed in three phases. The first phase

involved burst and collapse tests and single-frequency eddy-current (EC) examinations of

typical steam generator tubing with precision machined flaws. The goal of Phase I was to

develop empirical models of remaining tube integrity as a function of flaw type and size,

and to determine the capability of EC inspection methods to detect and size tube degradation. In Phase II, a smaller number of specimens with the same flaw types were

investigated, but tube specimens were degraded by chemical means rather than machining

methods. This approach was used to better simulate the irregular geometry of serviceinduced degradation. In the final phase of the program, the retired-from-service Surry

2A Steam Generator was used as a test bed to investigate the reliability of inservice EC

inspection equipment, personnel, and procedures, and as a source of service-degraded

tubes for further validating the empirical equations of remaining tube integrity.

12. KEY WORDS/DESCRiPTORS (List words or phrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT

steam generators, inservice inspection reliability, eddy Unlimited

current testinq, tube intecrity, burst testing, collapse 14. SECURITYCLASSIFICATION

testing, samplinq plans, Surry (This Page)

Unclassified

(This Report)

Unclassified

15. NUMBER OF PAGES

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