ATOMIC ENERGY FSLWM L'ENERGIE ATOMIQUE OF CANADA … · atomic energy of canada limited "team...

AECL-7251

ATOMIC ENERGY FSLWM L'ENERGIE ATOMIQUEOF CANADA UMITED T ^ S j f DU CANADA LIMITEE

STEAM GENERATOR TUBE PERFORMANCE:EXPERIENCE WITH WATER-COOLED NUCLEAR

POWER REACTORS DURING 1979

Performance des tubes de generateur de vapeur:experience acquise en 1979 avec les reacteurs de puissance

refroidis par eau

O.S. TATONE and R.S. PATHANIA

Chalk River Nuclear Laboratories Laboratoires nucle'aires de Chalk River

Chalk River, Ontario

March 1981 mars

ATOMIC ENERGY OF CANADA LIMITED

"TEAM GENERATOR TUBE PERFORMANCE:

EXPERIENCE WITH WATER-COOLED NUCLEAR

POWER REACTORS DURING 1 9 7 9

by

O.S. Tatone and R.S. Pathania

Chalk River Nuclear LaboratoriesChalk River, Ontario

1981 March

AECL-7251

L'ENERGIE ATOMIQUE DO CANADA, LIMITEE

Performance des tubes de générateur de vapeur:expérience acquise en 1979 avec les réacteurs de puissance

refroidis par eau

par

O.S. Tatone et R.S. Pathania

Résumé

On a passé en revue la performance, en 1979, des tubes de générateurde vapeur dans les réacteurs de puissance refroidis par eau. Des défail-lances de tubes se sont produites dans 38 des 93 réacteurs étudiés. Ondécrit les causes de ces défaillances et les procédures conçues pour yremédier. Le taux des défaillances a été deux fois plus élevé en 1979 qu'en1978 mais il s'est avéré inférieur à celui des deux années précédentes. Lesméthodes employées pour détecter les défauts comprennent l'emploi accru desessais par courants de Foucault en multifréquençes et une tendance à 1'inspec-tion en pleine longueur de tous les tries. Pour réduire l'incidence desdéfaillances de tubes par la corrosion, les exploitants des centrales ontrecours à la déminéralisation du condensât en plein débit et à des tubes decondenseur plus étanches.

Laboratoires nucléaires de Chalk RiverChalk River, Ontario

Mars 1981

AECL-7251

ATOMIC ENERGY OF CANADA LIMITED

STEAM GENERATOR TUBE PERFORMANCE: EXPERIENCEWITH WATER-COOLED NUCLEAR POWER REACTORS DURING 1979

by

O.S. Tatone and R.S. Pathania

ABSTRACT

The performance of steam gf:;-";or tubes in water-cooled nuclear powerreactors has been reviewed for 1979. Tube failures occurred at 38 of the93 reactors surveyed. Causes of these failures and procedures designed todeal with them are described. The defect rr.te was twice that in 1978 butstill lower than the two previous years. Methods being employed to detectdefects include increasing use of multifrequency eddy-current testing and atrend to full-length inspection of all tubes. To reduce the incidence oftube failures by corrosion, plant operators are turning to full-flow condensatedemineralization and more leak-resistant condenser tubes.

Chalk River Nuclear LaboratoriesChalk River, Ontario

1981 March

AECL-7251

CONTENTS

1. INTRODUCTION 1

2. SURVEY OF 1979 FAILURES 2

3. HISTORY OF TUBE DEFECTS 19

4. CAUSES OF 19 79 TUBE DEFECTS 22

5. LOCATION OF 1979 TUBE DEFECTS 26

6. SECONDARY WATER CHEMISTRY CONTROL 26

7. STEAM GENERATOR TUBE MATERIALS 29

8. INSPECTION AND REPAIR PROCEDURES 31

9. SUMMARY 32

10. ACKNOWLEDGEMENTS 32

11. REFERENCES 33

APPENDIX A: STEAM GENERATOR DESIGN DATA 35

APPENDIX B: CUMULATIVE STEAM GENERATOR 41EXPERIENCE TO 1979 DECEMBER 31

TABLES

1. SUMMARY OF STEAM GENERATOR TUBES PLUGGED 3DURING 1979

2. TUBE DEFECTS VS YEAR 20

3. CUMULATIVE TUBE DEFECTS vs EFPD TO 211979 DECEMBER 31

4. 1979 TUBE DEFECTS vs EFPD 21

5. CAUSES OF 1979 TUBE DEFECTS 23

6. LOCATION OF 1979 TUBE DEFECTS 27

7. SECONDARY WATER CHEMISTRY vs CORROSION DEFECTS 2 8IN 1979

8. EXPERIENCE WITH STEAM GENERATOR TUBE MATERIALS 30TO 1979 DECEMBER 31

INTRODUCTION

Steam generators are large tube-.i.n-shell heat exchangers

in nuclear power plants where heat generated in the core is trans-

ferred to the secondary coolant to raise steam. Next to the reactor

vessel itself they are the largest components in a nuclear steam

supply system. Several thousand relatively thin-walled tubes

(-1.2 mm) separate the primary reactor coolant (in tubes) from

the secondary or steam side (in shell). During boiling, non-

volatile impurities in the feedwater can concentrate in stagnant

areas such as crevices on the secondary side providing conditions

conducive to tube corrosion.

Since a single tube leak may require a plant shutdown

there is incentive for high reliability in steam generators. Costs

associated w? th steam generator repair and with compulsory inspec-

tions to ensure tube integrity include fossil-fired replacement

power and radiation exposure to maintainers and inspectors. An

understanding of tube failure mechanisms and development of methods

to prevent or mitigate them can ensure higher reliability than has

been experienced in the past.

Experience with steam generator tubes in water-cooled1—8

reactors during 1979 has been surveyed as in previous years

The survey was conducted by questionnaires mailed directly to

station operating staff supplemented by a literature survey. Water-

cooled power reactors of capacity greater than 50 MW(e) (except

for NPD*) and with a minimum of 100 effective full-power days

(EFPD) of operating experience at 1979 December 31 have been included.

Reactors in Eastern Bloc countries have been excluded because of

lack of data.

Appendix A lists design parameters pertaining to steam

generator tube performance for 95 reactors and Appendix S lists

their cumulative experience.

*NPD - Nuclear Power Demonstration Reactor, Rolphton, Ontario, 25 MW(e)

- 2 -

Reactors surveyed in 1979 were of the following types:

74 pressurized water reactors (PWR)

14 pressurized heavy water reactors (PHWK)

4 boiling water reactors (BWR)

1 water-cooled, graphite reactor (LGR)

Two boiling water reactors with steam generators, KWL Lingen and

KRB Gundremmingen, are included in the Appendices for completeness

but these are now permanently shutdown. Seven PWR's and one PHWR

have been added since the 1978 survey.

SURVEY OF 1979 FAILURES

Experience at power reactors in which steam generator

tubes failed during 1979 is described below and summarized in

Table 1.

BEZNAU-L "2, SWITZERLAND

Fifteen tubes were plugged in the steam generators of

Unit-1, 8 because of phosphate wastage just above the tubesheet,

6 because of stress-corrosion cracking within the tubesheet crevice

and 1 because of fretting at the U-bend. The number of wastage

defects, resulting from phosphate treatment of secondary water

between 1971 and 1974, has diminished since the change to all-

volatile treatment (AVT). Stress-corrosion cracking in the

tubesheet crevice caused failures during AVT chemistry control in

1969-70 and after 1974 and also during the period of phosphate

additions. The steam generators at Beznau are similar to that at

Jose Cabrera which has had fretting defects (6 tubes) at the U-bend.

The steam generators at KWO Obrigheim are also similar and U-bend

defects have occurred there but by what appears to be a cracking

mechanism.

Table 1 - SUMMARY OF STEAM GENERATOR Tt'BES PLUGGED DURING 1979 '(Abbreviations found at end of table)

REACTOR

BeznaJ-1

Beznau-2

Borssele

Bruce-2

Bugey-3

Crystal River-3

Doel-2

Farley-1

Ginna

Indian Point-2

Indian Polnt-3

Jose Cabrera (Zorita)

KKS Stade

KWO olrighein

Hlhams-1

Mlhama-2

MIllstone-2

North Anna-1

TUBESPLUGGED

15

2

51

2

6

16

118

1

19

26

437

4

2

15

2

25

4

284

FAILURE

CAUSE

6 SCC*8 wastage1 fretting

1 SCC1 UD

50 wastage

UD

UD

mechanicaldamage

mechanicaldamage

42 SCC76 SCC

UD

13 SCC4 wastage2 UD

denting

denting

fretting1 *;astage

some wastage

5 SCC10 OT>

-

SCC

3 denting

denting

FAILURELOCATION

TS creviceabove TSU-bend

TS crevice

UD

above TS

U-bend, TSP

above TS

seal weld

TS crevice

U-bend

U-bend

TS creviceabove TSTSP

TSP

TSP, U-bend

AVB1 above TS

above TS

TS2 TSP, 1 U-bend,7 UD

-

TS crevice

TSP, 1 seal i

U-bend

, Id

SECONDARY

CHEMISTRY

CONTROL

AVT

AVT

TO4

AVT

AVT

AVT/CD

AVT/CD

AVT

AVT/CD

AVT

AVT

P°4

P°4

AVT

AVT

AVT

AVT

AVT/CD

CONDENSER

COOLING

WATER

fresh

fresh

sea

fresh

"resh

sea

brackish

fresh

fresh

brackish

brackish

fresh

fresh

fresh

sea

sea

sea

fresh

CONDENSER

LEAKS

no

no

yes

no

yes

yes

yes

no

no

ves

no

no

yes

yes

no

yes

yes

yes

COMMENTS

both leaking

2 removed for examination

both leaking

OTSG

35 leakers in TS crevice1 leaker at U-bend

tubes removedfor examination

13 leaking

removed for examination

2 leaking2 removed

1 leaking

2 leaking

Table 1 - conc'd

REACTOR TUBESPLUGGED

FAILURECAUSE

FAILURELOCATION

NPD

N-Reactor

Oconee-1

Oconee-3

Pelisades

Point Beach-1

Point Beath-2

Prairie Island-1

Klnghals-2

Robinson-2

Salem-1

San Onofre-1

SENA (Chooz)

St. Lucie-1

Surry-1

Three Mile Islaad-1

Tihange-1

37

50

86

2

23

1

6

67

39

3

27

wastage

75 erosion,3 fretting,,8 error

fretting

11 wastage6 denting6 UD

sec

wastage

fretting byforeign object

denting

u sec23 wastage5 UD

mechanicaldamage

sec3 denting

fretting

maintenancedamage

denting

ID

sec

TSP,I U-bend

TS crevice,5 above TS

above TS

near TS

U-bend, 1 TSP

TS creviceabove TS3 TSP, 2 UD

above TS

TSTSP

AVB

neat drilled TSP

2 TS crevice26 TSP

UD

above TS

SECONDASY CONDENSER CONDENSER COMMENTSCHEMISTRY COOLING LEAKSCONTROL WATER

PO.

AVT,partial CD

AVT/CD

AVT/CD

AVT/CD

AVT

AVT

AVT

PO,

yes

AVT

AVT

AVT/CD

AVT/CD

AVT

fresh

fresh

fresh

fresh yes

fresh yes

Sresh

fresh

sea

fresh

brackish

no

NR

yeB

yes

yes

4 leaking, horizontalsteam generator

SS tubes only, horizontalsteam generators

1 tube with large leak

2 leaking

2 leaking

1 leaking

fresh

fresh

NR

yes

SECONDARY CONDENSER CONDENSER COMMENTSCHEMISTRY COOLING LEAKSCONTROL WATER

AVT fresh

AVT/CD fresh

AVT sea

no

AVT

Table 1 - cont'd

TUBES

PLUGGED

FAILURE

CAUSE

FAILURE

LOCATION

Trlno Vercellese

Trojan

Turkey Polnt-3

Turkey Point-4

1

9

740

185

fretting

LT)

denting

40 wastage

denting6 wastage

U-bend

U-bend

TSP

above TS

TSPabove TS

ABBREVIATIONS USED I» TABLE 1

AVB - antivibration bar

AVT - all-volatile treatment

CD - condensate demineralization

IGA - intergranular attack

NR - not reported

OTSG - once-through steam generator

FO, - phosphate treatment

SCC - stress-corrosion cracking

SS - stainless steel

TS - tubesheet

TSP - tube support plate

UD - undetermined

- 6 -

Two steam generator tubes were plugged at Beznau-2. One

defect was caused by stress-corrosion cracking within the tube-

sheet crevice, while location and cause of the other failure were

not determined. Beznau-2 had phosphate treatment of secondary

water between criticality in 1971 and 1974. Stress-corrosion

cracking in the tubesheet crevice was not observed until 1977 and

1978 (1 tube each year).

Inspection of tubes was carried out by automated eddy-

current equipment. In Unit-2, 16 tubes were inspected to the first

support in steam generator A. In Unit-1 a more extensive inspection

of both steam generators consisted of testing 3193 tubes to the

first support plate, 737 tubes to the top support plate and 199

tubes around the U-bends.

BORSSELE, NETHERLANDS

Fifty-one tubes were plugged at Borssele during 1979

(31 in steam generator #1; 20 in #2). Fifty of these defects were

caused by phosphate wastage just below the surface of the sludge

deposit on the tubesheet. Low phosphate treatment (2-6 mg-kg ;

Na:P04 =2.0) of secondary water has been practised at Borssele

since criticality in 1973. The steam generators are tubed with

Alloy-800. This is the first operational experience of significant

phosphate wastage in Alloy-800 tubes subject to low phosphate water

treatment.

Both steam generators were inspected by automated eddy-

current methods. In steam generator #1, 22 30 tubes were inspected

in the hot leg and 205 tubes were inspected through the cold leg

and U-bend. In steam generator #2, 1650 tubes were inspected in

the hot leg and 106 were inspected through the cold leg and U-bend.

All tubes with 50% wall thinning were explosively plugged. Two

tubes were removed from steam generator #1 for metallurgical tests.

Sludge has accumulated in the low-flow ("banana") zone of the tube-

sheet to a maximum height of 12.5 cm. No attempt was made to

reduce this deposit.

- 7 -

BRUCE-2, CANADA

Two tubes were plugged in steam generator #6 of Bruce-2.

One tube was leaking just above the point where the U-bend begins.

The other defect located at the top tube support plate was found

by eddy-current inspection. This is the second instance of tubes

failing at the U-bend tangent. In 1977 several tubes in steam

generator #3 had similar defects. The cause of these defects is

not known.

BUGEY-3, FRANCE

Six steam generator tubes failed in Bugey-3. In steam

generator 2, three tubes were plugged because of wall thinning of

10-60%. These were caused by a foreign object. In steam generator-3,

one tube leaked and two others were found to have wall thinning of

about 45%. These were also caused by a foreign object and were

located about 10 cm above the tubesheet.

Inspection methods used at Bugey include ultrasonics,

radiography, visual and eddy-current. All steam generators had a

complete baseline inspection prior to initial criticality.

CRYSTAL RIVER-3/ USA

Sixteen tubes were plugged because of mechanical damage

at the seal well's caused by debris from a burnable poison rod

assembly which worked loose and broke up in 1978 February. Seven

tubes had been plugged in 1978 because of this event.

DOEL-2, BELGIUM

At Doel-2, 118 steam generator tubes were plugged. Forty-

two tubes (35 leaking) were plugged because of defects within the

tubesheet crevice. The defects were located at the roll-expanded

zone and were caused by stress-corrosion cracking. One tube was

plugged because of a large leak at the U-bend. The defect was a

longitudinal crack at one of the small-radius (row 1) tubes and was

thought to be caused by stress-corrosion cracking. All tubes in

row 1 were gauged and 75 tubes with ovality exceeding 10% were

plugged.

Inspection at Doel is performed by automated eddy-current

equipment at 3 frequencies. About 12% of the tubes were inspected

throughout their full length.

FARLEY-L USA

One leaking tube was plugged at Farley-1. Th<_; defect was

located at the U-bend but the cause of failure was not determined.

Eddy-current inspection was performed on 153 tubes in

steam generator A and 306 tubes in steam generator C where the

leaker was located. Remote television inspection was used to aug-

ment eddy-current testing.

GINNA/ USA

Nineteen tubes, all in steam generator B, were plugged at

Ginna during 1979. Thirteen of these had indications of inter-

granular attack in the tubesheet crevice- 2 tubes showed wall

thinning just above the tubesheet. Tube corrosion by intergranular

caustic stress-corrosion cracking is typical of steam generators

with a long tubesheet crevice. At Ginna these failures have occurred

every year since 1975, the year after introduction of AVT control

of secondary water chem. ; _ry. The wall thinning at support plates

#1 and #2 was thought to be caused by water flashing to steam in the

annulus during the early years of operation. The tube support plate

annuli are now packed with corrosion products. Other tubes have

this typp of defect but it is <20% of the tube wall thickness. The

wastage defects are thought to be caused by a hydraulic-mechanical

mechanism rather than corrosion because all tubes are in the

periphery of the bundle where sludge does not normally accumulate.

Tube inspection was performed by multifrequency eddy-

current testing as in 1977 and 1978. The pattern of tubes inspected

was similar to that of 1978. Most tubes were tested to the first

support plate, some to the sixth support plate and a few over the

U-bend. About 2000 tubes were tested in each steam generator with

a 5:1 ratio between hot and cold legs.

Ginna was the first PWR station with recirculating steam

generators to employ full-flow deep-bed condensate demi-idralizationg

in the United States . Very good experience has been reported with

steam generator chemistry control and with the operation of the

demineralizer system.

INDIAN POINT-2, -3, USA

Twenty-six steam generator tubes were plugged at Indian

Point-2 because of reduced tube diameter at the support plates.

These defects were found by eddy-current inspection of 1519 hot

leg tubes.

Denting, a phenomenon caused by ingress of chloride

leading to acid-forming conditions, results in non-protective

corrosion product deposition in tube-to-tube support annuli in

steam generators with drilled-hole carbon steel support plates.

It has been postulated that addition of boric acid to secondary

water mitigates denting by forming stable, protective iron borates.

Indian Point-2 is now using this treatment.

Four hundred and thirty-seven tubes were plugged in the

four steam generators of Indian Point-3. Denting defects were

observed in 69 tubes at support plate intersections. Because

- 10 -

denting causes inward support plate distortion giving rise to the

potential for stress-corrosion cracking at the small-radius U-bends,

all row 1 tubes were plugged (368 tubes).

The steam generators in Unit-3 were inspected by techniques

commonly used at plants with significant denting. This includes use

of eddy-current probes of different diameter and photography of the

secondary side to measure distortion of flow slots. The sludge

deposit on the tubesheet was found to be soft and it was estimated

that about 92% could be removed by water lancing. Boric acid is

added to steam generators during condenser leakage.

JOSE CABRERA/ SPAIN

Three tubes were plugged because of fretting at the anti-

vibration bars and 1 because of phosphate wastage just above the

tubesheet. Only 7 tubes have been plugged in the Jose Cabrera

steam generator in 2915 effective full-power days of operation

with phosphate treatment of secondary water and 6 of these were

caused by fretting at the antivibration bars.

Multifrequency eddy-current testing was used to inspect

80 tubes at the U-bend, and almost all tubes to the first support

plate. Phosphate wastage of 40-49% of tube wall was detected in

6 tubes (including 1 which was plugged) and wastage of 30-39% was

detected in 46 tubes. This is the first reported instance of

phosphate wastage.

KKS STADE, FRG.

Eddy current inspection of 574 tubes in steam generator

#1 and 1262 tubes in steam generator #2 showed that 3 tubes in

#1 and 56 in #2 had phosphate wastage of less than 25% of the tube

wall. Two tubes were removed for metallurgical examination. Stade,

like Borssele, has Alloy-800 tubes and has used low phosphate

treatment (2-6 ing "kg ; Na:P0^ = 2.6) since initial operation in

1972.

- 11 -

KWO OBRIGHEIM, FRG

Fifteen tubes were plugged at Obrigheim. Defects,

including 13 leaking tubes, were caused by secondary-side stress-

corrosion cracking just above the tubesheet {5 tubes) and by

undetermined mechanisms at the support plates (2 tubes) and the

U-bend (1 tube). The location of 7 other defects was not

determined.

Eddy-current inspection was performed on 455 hot-leg tubes

in steam generator # 1 and 1451 hot leg tubes in steam generator #2.

An additional 454 tubes in #2 were inspected throughout their full

length.

MIHAMA-2, JAPAN

Twenty-five tubes were plugged at Mihama-2 because of

stress-corrosion cracking in '.he tubesheet crevice. These are the

first tubes to be plugged at Mihama-2 because of this failure

mechanism and the first tubes to fail since 1975.

All tubes were inspected throughout their full length by

automated eddy-current methods. Full inspection of all tubes is

becoming standard practice at PWR's in Japan. During 1979 it was

done at all reactors except for the new Ohi-2 plant where only

25% of the tubes were inspected.

MILLSTONE-2, USA

Four tubes were plugged, including 1 which was thought

to have a leaking seal weld. The others were plugged because they

would not allow an eddy-current probe to pass.

- 12 -

Extensive eddy-current testing was performed in both

steam generators. Some denting was observed at the tubesheet and

at some tubes in the lower "egg-crate" (lattice bar) supports on

the hot side. The degree of tube constriction was very low with

a maximum of 0.05 mm at the tubesheet of steam generator #1

(0.04 in 1978). In the cold side, denting was observed on a few

tubes at the upper egg-crate supports but not at lower ones. All

tubes which pass through the partial, drilled-hole support plates

showed denting (0.3 mm average at #1 cold side, lower support plate)

Comparison with the 1978 inspection shows that both the number of

tubes affected and the degree of denting have increased slightly.

During inspection of the sludge deposit bands of high-conductivity

sludge were found above the usual (mainly magnetite) pile. This

condition was not present at the previous inspection.

NORTH ANNA-1/ USA

A multifrequency eddy-current inspection was performed

during the first refuelling and maintenance outage. About one-

third of the tubes inspected reacted to the 7.5 kHz probe indica-

ting corrosion of tube supports. In addition 2 dents were

discovered at tube support plates and 2 U-bends were leaking. The

cause of this damage was traced to ingress of resin from Powdex

demineralizers into the steam generators early in 1979 . At

steam generator operating conditions the resin decomposes to form

acid sulphates which behave similarly to acid chlorides in that

they can cause denting. To prevent future U-bend leaks all 282

first row tubes were plugged as well as the 2 tubes which displayed

denting at the tube support plate. The steam generators were

flushed and power-cycled to remove sulphates. Boric acid was added

to the steam generator for the next operating cycle.

- 13 -

NPD, CANADA

Thirty-seven tubes were plugged at NPD during 1979. In

contrast to earlier failures which were caused by fretting at the

tube support in a high flow region, the defects found in 1979 were

near the inlet tubesheet and were postulated to be caused by a

corrosion mechanism, probably phosphate wastage. This is the

first .Instance of corrosion defects in a CANDU steam generator.

The eddy-current inspection was performed manually (1200 tubes

to the first support plate, 50 full length). It is intended that

one or more tubes will be removed during 1981 for metallurgical

examination.

N-REACTOR, USA

N-Reactor has twelve horizontal steam generators of

which 10 are tubed with Alloy-600 and 2 (5A and 5B) are tubed with

type 304 stainless steel. Fifty tubes were plugged during 1979

in steam generators 5A and 5B. Thirty-four of these were leaking

and 16 were plugged on the basis of eddy-current testing.

The pattern of tube failures has been predictable at

N-Reactor. Originally there were 10 steam generators tubed with

stainless steel. Pre-operational and early operational failures

of the tubes by stress-corrosion cracking led to installation of

two additional steam generators with Alloy-600 tube 5 and progressive

retubing of all other steam generators except 5A and 5B with

Alloy-600. There have been no failures in the Alloy-600 tubes in

up to 2238 EFPD. Secondary water chemistry has been controlled

by all-volatile treatment and condensate demineralization has been

practised, but not full-time.

- 14 -

0C0NEE-.1, -3, USA

Eighty-six tubes were plugged in the Ooonee-1 steam

generators. Seventy-five of these defects were caused by an

erosion mechanism and 2 were caused by fretting, all at support

plate intersections. About one-third of the tubes were subjected

to eddy-current inspection throughout their entire length. Both

400 kHz and multifrequency eddy-current systems were used.

At Unit-3, 2 tubes were plugged because of fretting at the tube

supports and 5% of the tubes were inspected by single-frequency

eddy-current.

PALISADES, USA

Twenty-three steam generator tubes were plugged at

Palisades; 11 had degraded by phosphate wastage, 6 were dented and

6 showed multiple eddy-current indications which require plugging

at 30% penetration of the tube wall. All defects were located at

tube supports with the exception of 1 at the U-bend.

Eddy-current inspection was performed on 3344 tubes

from the hot leg through the U-bend to the upper cold leg support,

and on 895 cold leg tubes. Use of the full-flow condensate

demineralizer system was restricted because of carry-over of resin

fines into the steam generators.

POINT BEACH-.V -?.. USA

At Unit-1, 283 tubes were plugged because of corrosion

in the tubesheet crevice and 5 were plugged because of corrosion

just above the tubesheet. All defects were on the inlet leg of

the tubes. Three tubes were removed for further examination.

This appears to be the same type of problem as experienced at

several other plants (Beznau, Doel, Ginna, Mihama-2 and Robinson-2)

- 15 -

The tubes are roller-expanded at the primary side of the tubesheet

leaving a long crevice on the secondary side {up to 500 nun deep in

some plants). This crevice concentrates corrosive chemicals and

causes either stress-corrosion cracking or intergranular attack.

There were 4 shutdowns for Lube plugging in addition to the

refuelling and maintenance outage. It was apparent from later

inspections that corrosion was proceeding rapidly. Replacement

of the steam generators is being considered unless the failure

rate can be reduced . One tube was plugged at Point Beach-2

because of wastage located ?r>out 75 mm above the tubesheet. Multi-

frequency eddy-current testing was used at both units.

PRAIRIE ISLAND-1, USA

During October 1979 one tube in steam generator A

ruptured giving a leak rate greater than 18.9 dm -s *. The unit

was shut down and the leaking tube was located about 75 mm above

the tubesheet on the tube-bundle periphery. Fretting by a coil

spring (used in sludge lancing equipment) had caused a perforation

measuring about 35 mm axially and 12 mm circumferentially on one

tube and lesser damage on two other tubes. These tubes and 3

adjacent ones were plugged and both steam generators were inspected

for additional loose parts. Some were found, but these were

immobile and had caused no damage. Significant tube corrosion has

not been observed in the steam generators of either Prairie Island

unit in over 1400 effective full-power days of operation. These

plants have used all-volatile treatment of secondary water and

although condensate demineralizers are available they have not been

used full-time.

RINGHALS-2, SWEDEN

Three steam generator tubes were plugged because of

defects at the U-bend and 1 because of denting at a support plate.

* 300 gpm (US)

- 16 -

The remaining 63 tubes in row 1 (small radius U-bend) were also

plugged as a preventive measure. About 35% of the tubes were

inspected by automated eddy-current techniques and the support

plates were photographed. One-half of the condenser has been

retubed with titanium.

ROBINSON-2, USA

Robinson-2, one of the few plants using high phosphate

(10-80 ing.kg ; Na:PO4 = 2.3-2.6) treatment of secondary water,

continued to experience moderate wastage of steam generator tubes

in 1979 (23 plugged). Eleven tubes were plugged because of

stress-corrosion cracking in the 'tubesheet crevice, while three

defects located at support plates were caused by an unknown mechanism.

The total number of tubes plugged was 39. Tube damage by corrosion

cannot be considered severe as the above defects were located

after eddy-current inspection of more than 70% of the tubes.

SALEM-L USA

Ten tubes in each of the four steam generators were

plugged because of damage caused by tube-lane blocking devices.

These devices are baffles which were retrofitted to some steam

generators to direct water flow away from the open lane and into

the tube bundle and hence provide more effective clearing of

sludge. They have caused tube defects at Salem-1 and indirectly

at Prairie Island-1 when a coil spring lodged between tubesheet

and baffle and caused fretting of 3 tubes.

Some denting has been observed at Salem but steps taken

to mitigate it seem to have been effective. The condenser with

90-10 cupronickel tubes has had considerable leakage and has now

been retubed with a more corrosion resistant Pe-Ni-Cr-Mo alloy

(AL-6X). A condensate demineralization system was also installed

during the first refuelling outage.

- 17 -

SAN ONOFRE-1, USA

Twenty-one tubes were plugged in steam generator A

including 1 leaker and 3 showing distorted eddy-current signals

at the support plates. The other tubes had defects located above

the tu^sheet. These were suspected to be caused by stress-12corrosion cracking

SENA (CHOOZ), BELGIUM

Three steam generator tubes were plugged at SENA because

of fretting at the antivibration bars. This is the only tube-

failure mechanism which has been observed at SENA in over 2500

effective full-power days of operation with stainless steel tubes.

ST. LUCIE-1, USA

Four tubes were plugged in steam generator IB because

of damage incurred during maintenance. Regular in-service inspec-

tion of 894 tubes in steam generator 1A inlet and 900 tubes in

steam generator IB inlet showed no tube degradation. St. Lucie

has titanium condenser tubes and full-flow condensate demineral-

ization is being planned.

SURRY-1, USA

Twenty-eight steam generator tubes were plugged at

Surry-1. Two tubes, located in the tubesheet crevice, had defects

originating on the primary side while the other 26 were plugged to

prevent future failures.

All tubes in the three steam generators were inspected

throughout.

- IS -

THREE MILE ISLAND-1/ USA

During the 1979 February refuelling outage 24% of the

steam generator tubes were inspected. Two tubes with 68% penetra-

tion of the tube wail, thought to be manufacturing defects, were

plugged. A third tube was plugged after an unsuccessful attempt

was mace to extract it

TRINO, ITALY

One steam generator tube was plugged because of fretting

at the antivibration bars. Like SENA and Yankee Rowe, Trino has

stainless steel steam generator tubes which have shown good reli-

ability and excellent corrosion behaviour.

Condenser tube material is being changed from cupronickel

to stainless steel.

TROJAN, USA

Nine steam generator tubes, including 5 with leaks, were

plugged at Trojan. The defects were located at the tangent where

the tubes begin to curve to form the U-bend. The cause of these

defects has not been determined. Eddy-current testing showed no

evidence of denting and ball probes showed tube ovality to be14within acceptable limits . Trojan uses all-volatile treatment

of secondary water and powdered resin condensate demineralizers.

TURKEY POINT-3/ -'1/ USA

Forty tubes were plugged because of continuing phosphate

wastage in Unit-3 steam generators. An addi'::.->nal 690 tubes were

plugged because of denting at support plates. Over 80% of hot-leg

tubes and up to 40% of cold-leg tubes were inspected by manual

eddy-current methods. Condensate demineralizers are being installed

in Turkey Point-3.

- 19 -

In Unit-4, 6 tubes were plugged because of phosphate

wastage and 173 were plugged because of denting. Up to 44% of

the hot-leg tubes and 23% of cold-leg tubes were eddy-current

tested.

As of 1979 December, 17% of the steam generator tubes in

Turkey Point-3 and 19% of the tubes in Turkey Point-4 had been

plugged. By 1980 June all condenser tubes, previously 90-10

cupronickel, had been replaced by titanium. A steam generator

replacement program has been initiated

HISTORY OF TUBE DEFECTS

In this report a tube defect is defined as one which

required the tube to be taken out of service. The history of tube

defects over the period of 1971-79 is summarized in Table 2. In

1979, thirty-eight reactors, constituting 41% of those surveyed

developed tube defects. There was a two-fold increase in the

number and percentage of defective tubes in 1979 compared with 1978.

At the end of 1979 there were more than 1.3 million steam gener-

ator tubes in service in 9 3 reactors of which 1.4% had been plugged

because of defects. If more than 20% of the tubes in a steam

generator are plugged then either the steam generators have to be

replaced or the reactor has to be derated. For a reactor with a

40 year lifetime this situation can arise if the annual tube defect

rate (the last column in Table 2) exceeds 0.5%. Steam generators

have already been replaced or will soon be replaced in reactors

such as Shippingport, Surry-1 and -2 and Turkey Point-3 and -4

because of a large number of tube failures.

The relationship between effective full-power days in

service and the tube defect rate (cumulative) is shown in Table 3.

Table 4 shows a similar relationship for defects in 1979. As in

previous surveys the reactors with greater than 1000 EFPD

exhibited a higher tube failure rate than those with less than

1000 EPPD. Nevertheless there were 11 reactors which exceeded

1000 EFPD without a single tube defect. These were Atucha-1,

- 20 -

TABT.F. 7

TUBE DEFECTS vs YEAR

YEAR

1 9 7 1 a

1972

1973

1974

1975

1976

1977

1978

1979

Reactors

in Survey

34

36

48

59

62

68

79

86

93

% withwith Defects Defects

19

13

11

25

22

25

34

31

38

56

36

23

42

35

37

43

36

41

in Survey

337 808

364 691

553 883

764 566b

805 376b

881 397b

1 085 825b

1 201 162

1 314 973

Tubes

with Defects

1 305

1 066

3 942

1 990

1 671

3 763

4 355

1 252b

2 687

% withDefects

0.39

0.29

0.71

0.26b

0.21

0.43

0.40b

0.10

0.20

a Cumulative to the end of 1971b Amended value

- 21 -

TABLE 3

CUMULATIVE TUBE DEFECTS vs EFPD*TO 1979 DECEMBER 31

EFPD

< 500

500-1000

> 1000

in Survey

15

22

56

Reactors

with Defects

4

9

45

% withDefects

27

41

79

in

211

345

758

Survej"

490

195

282

with

3

16

Tubes

Defects

332

789

444

% with

• r

l . i

2.2

* Effective Full-Power Days

TABLE 4

1979 TUBE DEFECTS vs EFPD*

EFPD

< 500

500-1000

> 1000

in Survey

15

22

56

Reactors

with Defects

4

6

28

% withDefects

27

27

50

in Survey

211 496

345 195

758 282

Tubes

with Defects

332

472

1 883

% withDefects

0.16

0.14

0.25

* Effective Full- Power Days

- 22 -

Biblis A, Calvert Cliffs-1, Cook-1, Kewaunee, MZFR, Pickering-1,

-3 and -4 and Zion-1 and -2.

The tube failure rates of various reactors can be

characterized by a tube failure index obtained by dividing the

fraction of defective tubes by the effective full-power years.

Appendix B shows that this failure index ranges from 0 (for reactors

with no tube defects) to 0-13 (for a reactor with a large number

of tube defects). Values between 0.005 and 0.008 imply that at

current failure rates steam generators may have to be replaced before

the end of design life. Fifteen to 20% of reactors surveyed are

in this category.

CAUSES OF 1979 TUBE DEFECTS

The causes of defects in 1979 are summarized in Table 5.

Various forms of corrosion (e.g. denting, SCC and wastage) accounted

for 90% of the tube failures.

DENTING

Denting, the leading cause of tube failures since 1976,

caused 64.4% of tube defects in 1979. The number of tubes plugged

because of denting increased from 923 in 1978 to 1733 in 1979.

Many of these tubes, particularly those in the innermost row, were

plugged to prevent future failures at the U-bends because of strains

induced by denting.

STRESS-CORROSION CRACKING

There was a sharp increase in the incidence of stress-

corrosion cracking (SCC) , which was the second most prevalent cause

of tube failures in 1979. The number of tubes plugged because of

SCC from the secondary side jumped from 80 in 1978 to 512 in 1979.

Most of the SCC failures occurred in the earlier models of steam

generators (e.g. Beznau-1, Doel-2, Pt. Beach-1) where there is a

deep tube-to-tubesheet crevice on the secondary side. Aggressive

- 23 -

TABLE 5

CAUSES OF 1979 TUBE DEFECTS

Cause Number of ReactorsAffected

Number of TubeDefects

of TubeDefects

64

19

6

2

2

0

0

.4

.0

.8

.8

.7

.4

.3

.4

Denting

SCC

Wastage

Erosion

Mechanical Damage

Fretting

Plugging Error

Unknown

10

9

11

1

5

6

1

13

1 733

512

183

75

72

12

8

92

- 24 -

species such as sodium hydroxide can concentrate in this crevice

by alternate wetting and drying and induce stress-corrosion

cracking. In some boilers (e.g. Ginna, Tihange-1) intergranular

corrosion occurred within or just above the tube-to-tubesheet

crevice. This attack is also believed to be caused by concentrated

hydroxide solutions. In Doel-2 longitudinal SCC occurred at the top

of the U-bend probably because of high stresses due to excess (>10%)

ovality. The straining of the U-bends in the innermost row because

of denting resulted in primary-side SCC in some reactors. These

defects are included under denting rather than SCC in Table 5.

The actions taken to minimize SCC in new units include

thermal treatment of Alloy-600 at 705°C for 15 h to improve its

resistance to SCC , shot peening of tubes to induce compressive

stresses , improved methods of making tube-to-tubesheet joints and18better control of secondary-side water chemistry . Of the 9 reactors

with SCC defects in 1979, 5 used all-volatile treatment, 2 were on

phosphate treatment and 2 used all-volatile treatment with

demineralization.

WASTAGE

Phosphate wastage caused 183 tube defects in 1979, up

from 86 in 1978. Of the 11 reactors where wastage was reported only

5 were using phosphate treatment. The other 6 changed over from

phosphate to all-volatile treatment several years ago and of these

2 used all-volatile treatment with condensate demineralization.

In these reactors the wastage was probably caused by residual

phosphate in the sludge above the tubesheet. It is also possible

that some of the tubes with wastage were not inspected until

recently. The first reported defects in Alloy-800 steam generator

tubes occurred in Borssele and Stade and were caused by wastage

in the sludge above the tubesheet. Wastage was also detected in

Jose Cabrera (Zorita) and one tube was plugged because of it.

This is the first report of wastage in this reactor which has

operated for 2915 EFPD. Wastage is also suspected to be the

- 25 -

cause of tube defects in NPD which has operated for 3970 EFPD. It

appears that the rate of wastage in the sludge above the tubesheet

varies significantly from reactor to reactor being low for reactors

such as Zorita and higher for other reactors (e.g. Surry-1, Ginna,

San Onofre, Borssele). In the past, rapid phosphate wastage has

been observed under support straps at U-bends in some reactors

(e.g. Mihama-1, Palisades, Shippingport). It would be useful to

examine the reasons for these differences.

EROSION

Erosion at a support plate was reported to be the cause

of 75 tube defects in the once-through steam generators at Oconee-1.

MECHANICAL DAMAGE

Five reactors reported tube failures caused by various

forms of debris in the steam generators. For example at Crystal

River-3 debris from the failure (in 1978) of a burnable poison

rod assembly damaged the seal welds. In Prairie Island-1 a steel

coil spring left behind from the equipment used to remove sludge

from the tubesheet damaged 6 tubes and caused a large leak. In

Salem-1, 40 tubes suffered wear damage when devices used to block

the tube-free lanes came loose.

FRETTING

Fretting wear by flow-induced vibration at antivibration

bars at U-bends occurred in 4 reactors (Beznau-1, Zorita, Trino

and SENA). Fretting was also observed at tube support plates in

Oconee-1 and -3. The percentage of tube defects by fretting has

declined steadily from 2.6% in 1976 to 0.4% in 1979 primarily as

a result of design modifications to existing steam generators.

- 26 -

OTHER CAUSES

In one reactor (Oconee-1) 8 tubes were plugged by error.

The causes of defects of 94 tubes in 14 reactors could not be

determined.

LOCATION OF 1979 TUBE DEFECTS

Tube support plates and U-bends were the two most common

locations for tube failures, accounting for 72% of the defects

(Table 6). While most of these defects were caused by denting,

others were caused by fretting and erosion. The use of more

corrosion-resistant tube support materials and more 'open' tube-to-

tube support crevices should help to reduce the incidence of

denting at tube supports. Since the defects at U-bends were also

caused by dent-induced strains their incidence should also decrease

in the future.

The defects within the tubesheet occurred by SCC in the

earlier models of steam generators which have a deep tube-to-

tubesheet crevice on the secondary side. In the later designs,

where this crevice is closed by expansion of the tube, no problems

have been experienced. The defects in the sludge above the tube-

sheet occurred by wastage or SCC. Improved mechanical or chemical

methods to remove sludge are needed to minimize failures above the

tubesheet. Blowdown has been ineffective in preventing the build-

up of sludge.

SECONDARY WATER CHEMISTRY CONTROL

The water chemistry used by various reactors in 1979 is

shown in Appendix B. Table 7 shows the relationship between

secondary water chemistry control and corrosion defects from the

secondary side. In 1979, 59% of the reactors used all-volatile

treatment (AVT), 28% used all-volatile treatment with condensate

- 27 -

Location

Within Tubesheet

Near Tubesheet

Tube Support Plate

U-bend

Undetermined

TABLE

LOCATION OF 1979

Number of R e a c t o r sAffected

10

17

17

12

6

6

TUBE DEFECTS

Number of TubeDefects

400

279

1 123

812

73

% of TubeDefects

14.9

10.3

41.8

30.2

2.8

TABLE 7

SECONDARY WATER CHEMISTRY vs CORROSION DEFECTS* IN 1979

Water Chemistry Reactors

Number With DefectsX withDefects

Tubes

Number With DefectsX withDefects

Phosphate

All-volatile&

Condensate Demineralization

12

26

50

19

126 954

420 098

145

464

0.11

0.11

CO

I

All-volatile 55 13 24 767 921 1 819 0.24

Includes denting, phosphate wastage and SCC from the secondary side

- 29 -

demineralization and 13% used phosphate treatment. The correspon-

ding percentages for 1976 were 71%, 19% and 10%,respectively. There

appears to be a trend towards increasing use of full-flow condensate

demineralizers. The additives used with AVT include hydrazine and

ammonia, morpholine or cylohexylamine. Most operators use ammonia

Four reactors (Indian Point-2 and -3, Ko-Ri-1 and North Anna-1) are

using boric acid to prevent denting. Most reactors on phosphate

treatment use less than 10 mg POA/kq H^O.

It is evident from Table 7 that the tube defect rate

with AVT and condensate-demineralization or with phosphate treat-

ment was significantly less than that with AVT alone. However

the percentage of reactors with defects was the highest with

phosphate treatment and the lowest with AVT and condensate demin-

eralization. Data from 1978 and 1979 surveys indicate that the

incidence of corrosion defects was very low when AVT with condensate

demineralization was used. However it is interesting to note that

of the 11 reactors which have operated for more than 1000 EFPD

without a single tube defect 7 (64%) were on AVT, 2 (18%) on

phosphate treatment and 2 (18%) on AVT with condensate demineral-

ization. This suggests that all three methods of chemical control

are viable and can be made to work. It would be useful to examine

the reasons for the excellent performance of steam generators in

these reactors.

STEAM GENERATOR TUBE MATERIALS

Table 8 shows the experience with steam generator tube

materials to the end of 1979. Alloy-600 was the most widely used

material accounting for 74% of all the tubes in service. Monel-400,

Alloy-800 and austenitic stainless steel accounted for 13%, 8% and

5% of the tubes in service. The tube defect rate of Alloy-600 and

stainless steel was considerably higher than that of Monel-400 and

Ailoy-800. However some of the defect mechanisms (e.g. denting)

are not related to the tube material.

- 30 -

TABLE 8

EXPERIENCE WITH STEAM GENERATOR TUBE MATERIALS TO 1979 DECEMBER 31

TubeMaterial

StainlessSteel

Alloy-600

Monel-400

Alloy-800

Number ofReactors

9

69

8

8

Number ofTubes

71 922

974 232

167 700

101 119

Number ofTube Defects

612

19 565

335

53

% withDefects

0 . 9

2 . 0

0.2

0.05

FailureMechanism

SCC.W

SCC,W,D,Fr,F

SCC,Fr

W

SCC - Stress-Corrosion Cracking

W - Phosphate Wastage

D - Denting

Fr - Fretting

F - Fatigue

- 31 -

Appendix A lists the tube and tube support plate materials

used in the reactors surveyed. It also gives the condenser tube

materials where available. There is a trend towards the use of

more corrosion-resistant materials (e.g. titanium, AL-6X and

70-30 Cu-Ni) for condenser tubes.

INSPECTION AND REPAIR PROCEDURES

Fifty-five of the reactors surveyed inspected some or all

of their steam generator tubes. Automated eddy-current was the

most common method of inspection used. Some operators used a

multifrequency eddy-current technique to detect tube degradation,

tube support cracking and sludge build-up. Operators of some

reactors (Fessenheim-1, Genkai-1, Ko-Ri-1, Mihama-1, -2, -3,

Millstone-2, Takahma-2, Ohi-1) conducted full-length inspections

of all their steam generator tubes in 1979. However most operators

inspected a limited number of tubes at each shutdown. Since

corrosion damage is more likely to occur in the hot leg more tubes

were inspected in the hot leg than in the cold leg. Visual inspec-

tion was used at two reactors (Crystal River-3 and Loviisa-1). In

Crystal River-3 some tube-to-tubesheet seal welds were inspected

by helium leak and dye penetrant tests.

Condenser leaks 'were detected by measurement of sodium

concentration in the condensate and cation conductivity in the

blowdown (and occasionally in the condenser hot well). Leaking

condenser tubes were located by covering the tubesheet with plastic

film or foam under vacuum or by helium leak detection methods.

- 32 -

SUMMARY

Compared to 1978, there was a two-fold increase in the

number of steam generator tube defects in 1979. Denting was again

the leading cause of 'tube failures, followed by SCC and phosphate

wastage. There was a sharp increase in the incidence of stress-

corrosion cracking in the tubesheet region in reactors with a

long tube-to-tubesheet crevice. The use of volatile treatment with

condensate demineralization is becoming more widespread and reactors

employing this treatment showed a low incidence of tube defects

because of corrosion from the secondary side.

ACKNOWLEDGEMENTS

The authors wish to extend grateful thanks to station

operating staff who have taken time from their already busy schedules

to supply information for the survey. We also extend thanks to

Mrs. Rhea Fraser for capable assistance in preparing the manuscript.

- 33 -

REFERENCES

1. STEVENS-GUILLE, P.D., Steam Generator Tube Failures: A World Survey of Water-Cooled Nuclear Power Reaetors to the End of 1971. Atomic Energy of CanadaLimited, Report AECL-4449, (1973).

2. STEVENS-GUILLE, P.D., Steam Generator Tube Failures: World Experience inWater-cooled Nualear Power Reactors During 1972. Atomic Energy of CanadaLimited, Report AECL-4753 (1974/.

3. STEVENS-GUILLE, P.D. and HARE, M.G., Steam Generator Tube Failures: WorldExperience in Water-Cooled Nuclear Power Reactors in 1973. Atomic Energy ofCanada Limited, Report AECL-5013 (1975).

4. HARE, M.G., Steam Generator Tube Failures: World Experience in Water-CooledNuclear Power Reactors in 1974. Atomic Energy of Canada Limited, ReportAECL-5242 (1976).

5. HARE, M.G., Steam Generator Tube Failures: World Experience in Water-CooledNualear Power Reactors in 197S. Atonic Energy of Canada Limited, ReportAECL-5625 (1976) .

6. TATONE, O.S. and PA1HANIA, R.S., Steam Generator Tube Failures: Experiencewith Water-Cooled Nuclear Power Reactors During 1976. Atomic Energy ofCanada Limited, Report AECL-6095 (1978).

7. PATHANIA, R.S. and TATONE, O.S., Steam Generator Tube Performance: Experiencewith Water-Cooled Nuclear Power Reactors During 1977. Atomic Energy ofCanada Limited, Report AECL-6410 (1979).

8. TATONE, O.S. and PATHANIA, R.S., Steam Generator Tube Performance: Experiencewith Water-Cooled Nuclear Power Reactors During 1978. Atomic Energy ofCanada Limited, Report AECL-6852 (1980).

9. HARHAY, A.J., FILKINS, D.L. and GRITS, G.J., Deep Bed Condensate Polishing -Retrofit of Ginna After One Year Operating Experience. Paper presented atthe American Power Conference, 41st annual meeting, April, 1979.

10. Nuclear Power Experience, Vol. PWR-2, section V.D., p. 107, paragraph 251.

11. (a) Nucleonics Week, December 13, 1979, p. 12.(b) Nuclear News, January 1980, p. 91.(c) Nualear News, April 1980, p. 43.

12. GREFN, S.J. and PAINE, J.P.N., Paper presented at an American Nuclear SocietyIbpical Meeting on Materials Performance in Nualear Steam Generators,St. Petersburg, Florida, 1980 October.


- 34 -



16. AIREY, G.P., Effect of ̂ Processing Variables on the Caustic Stress CorrosionResistance of Inconel Alloy 600, CORROSION 36_, p. 9-17 (1980).

17. Schuktanz, G., Inspection Findings at U-tube Steam Generators of GermanPressurized Water Reactors, Kerntechnik 20, p. 205-13 (1978).

18. Renshaw, R.H., Replacement of Damaged Steam Generator Tubing in 600 WeCANBU Nuclear Plants, Paper presented at ANS Topical Meeting on MaterialsPerformance in Nuclear Steam Generators, October 6-9, 1980, St. Petersburg,Florida, U.S.A.

- 35 "

APPENDIX A

STEAM GENERATOR DESIGN DATA

APPENDIX A: DESIGH DATA RELEVANT TO STEAM GENERATOR TUBE

Reactor Name

Arkansas One-1

Atucha-1

Beaver Valley-1

Beznau-1

Be2nau-2

Biblis A

Biblis B

Borssele

Bruce-1

Bruce-2

Bruce-3

Bruce-4

Bugey-2

Bugey-3

Bugey-4

Calvert Cliffs-1

Calvert Cliffs-2

Cook-1

Cook-2

Crystal River-3

Davis-Besse-1

Doel-1

Doel-2

Douglas Point

Size

MW(e)

net

820

320

852

350

350

1 146

1 240

447

750

750

750

750

920

920

900

850

850

1 054

1 065

825

906

392

392

208

First

Commercial

Operation

1974/12

1974/06

1977/04

1969/09

1971/12

1975/03

1977/01

1973/10

1977/09

1977/0"

1978/02

1979/01

1979/02

1979/02

1979/07

1975/05

1977/04

1975/08

1978/06

1977/03

1977/11

1975/02

1975/11

1968/09

(! of

SG*

2

2

3

2

2

4

4

2

8

8

8

8

3

3

3

2

2

4

4

2

2

2

2

8

Tubes

per SG

15 531

3 945

3 388

604

2 604

4 060

4 021

4 234

4 200

4 200

4 200

4 200

3 388

3 388

3 388

8 519

8 519

3 388

3 388

15 457

15 457

3 260

3 260

1 950

AreaperSG(m2)

12 304

3 454

4 785

3 097

3 097

4 510

4 335

3 600

2 368

2 368

2 368

2 368

4 780

4 7S0

4 780

8 424

8 424

4 784

4 784

12 245

12 245

4 130

4 130

970

SO TubeMaterial

600

800

600

600

600

800

800

800

600

600

600

600

600

600

600

600

600

600

600

600

600

600

600

400

Support Plate

Material 6. Type

CS-broached

SS-lattice

CS-drilled

CS-drilled

CS-drllled

SS-lattice

SS-lattice

SS-lattice

CS-broached

CS-broached

CS-broached

CS-broached

CS-drilled

CS-drilled

CS-drilled

CS-egg crate

CS-egg crate

CS-drilled

CS-drilled

CS-broached

CS-broached

CS-drilled

CS-drilled

CS-drilled

Builder

BW

GHH

U

W

W

KWU/DBM

KWU

Balcke

BW (Can)

BW (Can)

BW (Can)

BW (Can)

Fram

Fram

Fram

CE

CE

W

W

BW

BW

CKI.

CKL

MLW

Condenser

Tube

Material

Admiralty

Admiralty

SS

70-30 CuZn

70-30 CuZn

Admiralty

Admiralty

70-30 CuNi

Admiralty

Admiralty

Admiralty

Admiralty

Admiralty

Admiralty

Admiralty

70-30 CuNi

70-30 CuNi

AaCu

AsCu

70-30 CuNi

304-SS

Al Brass

Al Brass

Admiralty

Comments

OTSG

PHWR

-

CANDU

CANDU

CANDU

CANDU

OTSG

OTSG

CANDU

APPENDIX A - cont'd

Reactor Name

Dresden-1

Farlev-1

Fessenheto-1

Fessenheii»-2

Fort Calhoun-1

Garigliano

Genkai-1

Ginna

031 Neckar

Goesgen

Haddam Neck(Conn. Yankee)

Ikata-1

Indian Point-2

Indian Point-3

Jose Cabrera(Zorlta)

KANUPP

Kewaunee

KKS Stade

KKU Unterwesser

Ko-Ri-1

KKB Gundremmlngen

KHL Lingen

SizeMW(e)net

200

829

890

890

457

150

529

490

855

920

575

538

864

965

153

126

540

630

1 230

597

237

256

FirstCommercialOperation

1960/08

1977/12

1977/12

1978/03

1974/06

1964/06

1975/10

1970/09

1976/10

1979/11

1968/01

1977/09

1974/07

1976/08

1969/08

1972/12

1974/06

1972/05

1979/10

1978/04

1967/04

1968/10

1 OESG

4

3

3

3

2

2

2

2

3

3

4

2

4

4

1

6

2

4

4

2

3

2

Tubesper SG

1 801

3 388

3 388

3 388

5 005

1 785

3 388

3 260

4 052

4 106

3 794

3 388

3 260

3 260

2 604

1 355

3 388

2 993

4 021

3 388

1 929

5 000

Areaper -SG(nT)

605

4 784

4 780

4 780

4 428

560

4 784

4 129

4 270

5 400

2 573

4 785

4 129

4 129

2 308

705

4 785

2 930

4 335

4 785

870

2 360

SG TubeMaterial

SS

600

600

600

600

400

600

600

800

800

600

600

600

600

600

400

600

800

800

600

SS

SS

Support PlateMaterial & Type

CS-drilled

CS-drilled

CS-drilled

SS-egg crate

CS-d rilled

CS-drilled

CS-drilled

SS-lattice

SS-lattice

CS-drilled

CS-drilled

CS-drilled

CS-drilled

CS-drilled

CS-lattice

CS-driUed

SS-lattice

SS-lattice

CS-drilled

SS-lattlce

CS-drilled(SS-plated)

Builder

FW

W

Fram

Fram

CE

KM

MH1

W

GHH/Balcke

KWU/GHH

W

MHI

W

W

H

BH (Can)

W

DBW

KWU

W

VKW

GHH

CondenserTubeMaterial

Admiralty

Admiralty

Admiralty

304-SS

Al Brass

Admiralty

Admiralty

Admiralty

Admiralty

Al Brass

Admiralty

Admiralty

Admiralty

Al Brass

Admiralty

Admiralty

Admiralty

Al Brass

Comments

BWE

BWR

condenser beingretubed with 90-10CuNi

CANDU

BWR, shutdown

BWR, shutdown

APPENDIX A. - cont'd

Reactor Name

KHO obrigheim

Loviisa-1

Haine Yankee

Mihama-1

Mlhama-2

Mihama-3

Millatone-2

MZFR

North Anna-1

NPD

N-Reactor

Oconee-1

Oconee-2

Oconee-3

Ohi-1

Ohl-2

Palisades

Pickerlng-1

Pickering-2

Pickering-3

SizeMW(e)net

328

liltQ

790

320

470

780

796

52

943

22

860

871

871

S71

1 120

1 120

700

514

514

514

FirstCommercialOperation

1969/03

1977/05

1972/12

1970/11

1972/07

1976/12

1975/12

1966/12

1978/06

1962/03

1966/07

1973/07

1974/09

1974/12

1979/03

1979/12

1972/03

1971/07

1971/12

1972/06

II ofSG

2

6

3

2

2

3

2

2

3

1

102

2

2

2

4

4

2

12

12

12

Tubes per SG Tube

per SG SG(m ) MaterialSupport PlateMaterial & Type


2 605 2 750 600

5 536 2 510 SS

5 703 5 405

4 426

765

3 388

2 069

1 920

15 531

15 531

15 531

3 388

3 388

8 519

3 381

920

4 785

577

1 486

12 304

12 304

12 304

4 785

4 785

7 368

600

600

3 260 4 130 600

3 388 4 785 600

8 519 8 424 600

SS

600

600

600SS

600

600

600

600

600

600

2 600 1 858 400

2 600 1 858 400

2 600 1 858 400

CS-lattice(SS plated)

CS-egg crateCS-drilled (2)

CS-drilled

CS-egg crate

CS-drilled

CS-drilled

CS-egg crateCS-drilled (2)

CS-drilled

CS-drilled

CS-straps

CS-broached

CS-breached

CS-broached

CS-drilled

CS-drilled

CS-drilled (14)

CS-egg crate (2)

CS-lattice

CS-lattice

CS-lattice

GHH/Balcke Admiralty

AEE 70-30 CuNi

50% SS,50% Al Bronze

CE

BW

BW

BW

W

MHI

CE

BW (Can)

BW (Can)

BW (Can)

Al Brass

MHI Al Brass

MHI Al Briss

CE 70-jj CuNi

GHH/Balcke Admiralty

W 304-SS

BW (Can) Al Brass

CE Admiralty

304-SS

304-SS

304-SS

Al Brass

Al Brass

90-10 CuNl

Admiralty

Admiralty

Admiralty

horizontal SG

condensers beingretubed withSS (AL6X)

CANDU, horizontal SG

LGR, horizontal SG

OTSG

OTSG

OTSG

OS

I

CANDU

CANDU

CANDU

APPENDIX A - cont'd

SizeMW(e)

Reactor Name net

FirstCommercial // ofOperation SG

Tubesper SG SG(m

SG Tube Support PlateMaterial Material & Type Builder


Pickering-4

Point Beach-1

Point Beach-2

Prairie Island-1

Prairie Island-2

Rancho Seco

RAPP-1

Ringhals-2

Robinson-2

Saleo-1

San Onofre

SENA (Chooz)

Shippingport

St. Lucie-1

Surry-1

Surry-2

Takahama-1

Takahama- -;

Tarapur-1

Tarapur-2

514

497

497

520

520

913

207

822

700

1 090

1973/06 12

1970/12 2

1972/10 2

1973/12 2

1974/12 2

1975/04 2

1973/12 8

1975/04 3

1971/03

1977/06

430 1968/01

280 1967/04

100 1957/12

802 1976/12

788 1972/12

788 1973/05

780 1974/11

780 1975/11

198 1969/10

198 1969/10

2 600

3 260

3 260

3 388

3 388

15 457

1 950

3 388

3 260

3 388

3 794

1 662

1 6923 050

8 485

3 388

3 388

3 388

3 388

1 589

1 589

1 858

4 129

4 129

4 786

4 786

12 245

970

4 784

4 128

4 784

2 573

1 385

1 2441 084

4 784

4 784

4 785

4 785

400

600

600

600

600

600

400

600

600

600

600

SS

600600

600

600

600

600

600

SS

SS

CS-lattice

CS-drilled

CS-drilled

CS-drilled

CS-drilled

CS-broached

CS-drilled

CS-drilled

CS-drilled

CS-drilled

CS-drilled

CS-drilled

CS-drilledCS-strap

CS-egg crat

CS-drilled

CS-drilled

CS-drilled

CS-drilled

CS-drilled

CS-drilled

BW (Can)

U

W

U

W

BW

MLW

W

W

W

W

CKL

BUFW

CE

W

W

W

MHI

Admiralty

Admiralty

Admiralty

SS

SS

SS

Admiialcy

70-30 CuNi

Admiralty

90-10 CuNi

50% CuNi,50% Ti

Admiralty

SS

Ti

90-10 CuNi

90-10 CuNi

Al Brass

Al Brass

FW

FW

OTSG

CANDU

condenser beingretubed with Ti

condenser beingretubed with AL6X


condenser beingretubed with CuNi

horizontal SG

condenserretubed fromAl Brass


condenser beingr tubed with Ti

BUR

BUR

APPENDIX A - c o n t ' d

Reactor Name

Three MileIsland-1

Tihange-1

Trino Vercellese

Trojan

Turkey Foint-3

SizeMk'(c)net

792

880

242

1 130

693

First

Operation

1974/09

1975/10

1965/01

19 76/05

1972/12

Area Condenserlubes \n--r ., SG Tubt> Suppor t P l a t e Tubepe r SG Sr;(m"> >lazpria]_ Material f> T̂ vpe Buildt'r^_ Material Ctimmeius _

15 511 12 014 600 CS-broached RW SS OTSC-

3 383 ', 788 600 CS-drilled CK1. Admiralty

1 662 1 384 SS CS-drilled VC CuNi condenser being

retubed with SS

3 3B8 a 785 600 CS-dri l led ;;

3 260 , 128 600 CS-dril U-J W

Turkey Point-4 693 1973/09 ; 3 260 '. 12S 600 CS-dril K-J

Yankee Rowe 175 1961/07 A 1 620 1 2iS SS

2ion-l 1 050 1973/12 4 3 ?60 4 128 600 CS-dri l l i -d

Zion-2 1 050 1974/09 4 1 260 4 128 600 CS-iiri 1 l « l

Adn

Ti

T ii : u -

Adt

SS

SS

(

ni

r a l t v

75%);(25*)

r a l t y

eondens.f rom Cu

i-ondensrt-tubed

c rS i

cru

retubed

beinsi t h Ti

ABBREVIATIONS USED TN APPENDIX A STFAM GENERÛOR MANLTACTl'RCRS

70-30 CuNi 70-30 Cupro-Nickcl70-30 Cu2n 70-30 Cupro-Zinc90-10 CuNi 90-10 Cupro-N'irkel304-SS Type 30i S t a i n l e s s Stc-ei400 Monel-400600 Alloy-600800 Alloy-800Admiralty Admiralty Brass (28/.rlSnCu nominal)AL6X High-Alloy S t a i n l e s s Steel fFe-M-Cr-Mu iAl Brass .Aluminum Brass (22ZruAlCu nonin.il)Al Bron2e Aluminum BronzeAsCu Arsenical Copper C>1T As)BWR Boiling Water ReactorCANDU Canada Deu te r ium Urar iumCS Carbon S t e e lLGR Lighc-watier-cooled pranhite rt'ictorOTSG Once-Through Steam G e n e r a t o rPHUR Pressurized Heavy Water ReactorSG Steam Generator

AF.i:Balcr-u•iv:H* (Can)ci-:CKL.DWBTrnir-izOr)HKMKUTMHI!>&'

W

AtiJmunerôexportBalcki'fl̂ hf-r.i-V K. Wi 1 ,-ovHabrock & Wilenx CanadaCorabnsrinn EngineeringCockiirll l[leutsche Bab.-nck i, WilcoxFr;imatomoFoster Whêlt-rCutenhof fnunpalmtti;Roninklizke Machincfabrik SKraftuerk VnionMitsubishi Heavy Industr i t i^Montreal Locomotive WorksVert'inpte KesselwcrkeWestinghoust'

- 41 -

APPFNDTX B

CUMULATIVE STEAM GENERATOR EXPERIENCETO 1979 DECEMBER 31

APPENDIX B: CUMMULATIVE STEAM GENERA'. 3R EXPERIENCE TO 1979 DECEMBER 31 *(Abbreviations at End of Tahle)

Reactor Name

Arkansas One-1

Atucha

Beaver Valley-1

Beznau-1

Beznau-2

Biblis A

Biblis B

Borssele

Bruce-1

Bruce-2

Bruce-3

Bruce-4

Bugey-2

Bugey-3

Bugey-4

Calvert Cliff.s-1

Calvert Cli££s-2

Cook-1

Cook-2

Crystal River-3

Davis-Besse-1

Doel-1

Size

.let

820

320

852

350

350

1 146

1 240

447

750

750

750

750

920

920

900

850

850

1 054

1 065

825

906

392

Number ofSG* Tubes

31 062

7 890

10 164

5 208

5 208

16 240

16 084

8 468

33 600

33 600

33 600

33 600

10 164

10 164

10 164

17 038

17 038

13 552

13 552

30 914

30 914

6 520

SecondaryChemistryControl

AVT/CD

P04

AVT

AVT

AVT

P04

4

P04

AVT

AVT

AVT

AVT

AVT

AVT

AVT

AVT/CD

AVT/CD

AVT

AVT

AVT/CD

AVT/CD

AVT/CD

PreviousChemistrvControl (Dateof Change)

-

-

-

PO,(74/07)4

P0.(74/09)4

-

-

-

-

-

-

-

-

-

-

-

-

-

-

-

P0,(74/10)

CondenserCool ingWater

F

F

F

F

F

F

F

S

F

F

F

F

F

F

F

B

B

F

F

S

F

B

EFPD

1 142

1 554

374

3 689

2 486

1 245

705

1 789

748

732

577

291

234

178

165

1 216

838

1 116

373

537

303

1 476

CumulatedTube Defects

5

0

0

1 007

275

0

0

51

0

11

0

0

0

6

0

0

0

0

0

24

0

24

Failuresper TubeYear (x 10 )

0.5

0

0

191.3

77.5

0

0

12.3

0

1.6

0

0

0

12.1

0

0

0

0

0

5.3

0

9.1

Comments

OTSG

PHUR

4 tubes removed forexamination

3 tubes removed forexamination

2 tubes removed forexamination prior t1979

CANDU

CANDU

CANDU

CANDU

OTSG

OTSG

AFPEKDIX B - cont'd

Reactor Name

Doel-2

Douglas Point

Dresden-1

Farley-1

Fessenheinj-1

Fessenheim-2

Fort Calhoun

Garigliano

Genkal-1

Ginna

GOT Neckar

Goesgen

Haddam Neck(Conn. Yankee)

Ikata-1

Indian Point-2

Indian Polnt-3

Jose Cabrera(Zorlta)

KSHUFP

Kewaunee

KKS Stade

KKD Untenresser

SizeMW(e)net

392

208

200

829

890

890

457

150

529

490

855

920

575

538

864

965

153

126

540

630

1 230

Number ofSG Tubes

6 520

15 600

7 204

10 164

10 164

10 164

10 010

3 570

6 776

6 520

12 063

12 318

15 176

6 776

13 040

13 040

2 604

8 130

6 776

11 972

16 084


AVT/CD

AVT

CD

AVT

AVT

AVT

AVT

-

AVT

AVT/CD

P04

P04

AVT

AVT/CD

AVT

AVT

TO4

AVT

AVT

TO4

TO,

PreviousChemistryControl (Dateof change)

PO4(75/02)

P04(74/U)

-

-

-

-

-

-

-

P04(74/ll)

-

-

PO4(75/O2)

-

PC4(75/02)

-

-

-

PO4(74/10)

-

_

CondenserCoolingWater

B

F

F

F

F

F

F

F

S

F

F

F

F

S

B

B

F

S

F

F

F

EFPD

1 184

2 181

3 256

448

536

489

1 530

3 249

1 184

2 439

853

156

3 416

628

1 206

815

2 915

886

1 530

2 414

303


147

2

180

2

0

0

3

332+

1

205

0

0

32

0

71

562

7

0

0

2

0

Failuresper Tube .Year (x 10 )

69.5

0.2

28.0

1.6

0

0

0.7

104.5

0.4

47.0

0

0

2.2

0

16.5

193.0

3.4

0

0

0.2

0

Comments

CANDU

BUR

BUR

boric acid added tosecondary water


CANDU

APPENDIX B - cont'd

Reactor Name

Ko-Ri-1

KRB Gundrenssingen

KWL Lingen

KHO Obrigheim

Loviisa-1

Maine Yankee

Mihama-1

Mlhama-2

Mihams-3

Millstone-2

HZFR

North Anna-1

NPD

"-Reactor

Oconee-1

Oconee-2

Oconee-3

OHI-1

OHI-2

Palisades

Pickerlng-1

Plckerlng-2

SizeMW(e)net

597

237

256

328

440

790

320

470

780

796

52

943

22

860

871

871

871

1 120

1 120

700

514

514

Number ofSG Tubes

6 776

5 787

10 000

5 210

33 216

17 109

•3 852

6 520

10 164

17 038

1 530

10 164

2 069

19 2003 840

31 062

31 062

31 062

13 552

13 552

17 038

31 200

31 200


AVT

AVT

neutral/CD

AVT

AVT

AVT

AVT/CD

AVT

AVT/CD

AVT/CD

PD4

AVT/CD

AVT/CD

AVT/CD

AVT/CD

AVT

AVT

AVT/CD

AVT

AVT

PreviousChemistryControl (Dateof Change)

-

-

-

-

-

PO^(75/09)

PO4(75/O9)

-

-

-

-

-

-

-

-

-

-

PO4(74/1O)

-

CondenserCoolingWater

S

F

F

F

B

B

S

S

s

s

F

F

F

F

F

F

F

S

s

F

F

F

EFPD

394

2 664

1 743

3 214

801

1 667

679

1 359

687

906

3 026

367

3 970

2 23S

1 426

1 250

1 205

225

159

1 271

2 554

2 442


0

364

112

273

0

15

2 208

297

0

766

0

284

47

092

161

12

19

0

0

3 696

0

1

Failuresper Tube ,Y ( in t

0

86.2

23.4

59.5

0

1.9

1 340.9

122.3

0

181.1

0

277.9

20.9

039.1

13.3

1.1

1.8

0

0

623.0

0

0

CD is planned, boricacid added tosecondary water

BWR, shutdown

BWR, shutdown

PHWR


CANDU, horizontal SG

CD not full-time,2 SG's tubed withstainless steel, LGR

OTSG

OTSG

OTSG

CD under construction


CANDU

CAIJDU

APPENDIX B - cont'd

Reactor Name

PreviousSize Secondary Chemistry CondenserMW(e) Number of Chemistry Control (Date Coolingnet SG Tubes Control ' of Change) Water EFPD

2 068

1 873

2 436

2 096

1 537

1 451

984

851

1 061

2 208

402

3 168

2 532

3 648

832

1 463

1 325

834

913

1 987

2 076


0

0

678

25

6

1

8

0

296

157

40

265

26

4260

4

1 798

1 924

196

0

4+

209

Failuresper Tube ,Year (x 10 )

0

0

155.8

6.7

2.1

0.4

1.0

0

100.9

26.5

26.8

26.8

5.6

125.90

1.0

441.3

521.5

84.4

0

2.3

115.6

Pickering-3

Pickeriag-4

Faint Beach-1

Point Beach-2

Prairie Island-1

Prairie Island-2

Rancho Seco

RAPP-1

Rlnghals-2

Robtason-2

Salem-1

San Onofre-1

SENA (Chooz)

Shlppingport

St. Lucie-1

Surry-1

Surry-2

Takahama-1

Takahama-2

Tarapur-1

Tarapur-2

Three MileIsland-1

514

514

497

497

520

520

913

207

822

700

1 090

430

280

100

802

788

788

780

780

198

198

792

31 200

31 200

6 520

6 520

6 776

6 776

30 914

15 600

10 164

9 780

13 552

11 382

6 648

3 3846 100

16 970

10 164

10 164

10 164

10 164

3 178

3 178

31 062

AVT

AVT

AVT

AVT

AVT/CD

AVT/CD

AVT/CD

AVT

AVT

F04

AVT/CD

P04

AVT

AVT

AVT

AVT/CD

AVT/CD

AVT

AVT

AVT/CD

PO, (79/12)

(74/12)

(74/Q9)

(74/10)

CANDU

CANDU

CD not full-time

CD not full-time

OTSG

CAHDU

CD being commissioned

horizontal SG - 2 0-tube, 2 straight-tube

CD planned

shutdown during 1979for SG replacement

BWR

BUR

APPENDIX B - cont'd

SizeMWCe)

Reactor Name

Tihange-l

Trino Vercellese

Trojan

Turkey Point-3

Turkey Point-4

Yankee Rowe

Zion-1

Zlon-2

net

880

242

1 130

693

693

175

1 050

1 050

SG

10

6

13

9

9

6

13

13

164

648

552

780

780

480

040

040


Previous

Chemistry

Control (Date

of Change)

CondenserCooling

Water F.FPDCumulatedlube Detects

Failuresper TubeYear (x 10 )

AVT

AVT

AVT/CD

AVT

AVT

AVT

AVT

AVT

(74/09)

(75/03)

1 192

3 D37

568

1 666

1 498

4 971

1 285

1 151

6

10

1 655

1 861

95

0

0

12.0

1.1

4.7

370.7

463.6

10.8

0

0

P0 4 used in

manufacturing



ABBREVIATIONS USED IN APPENDIX B

AVT8BWRCANDUCD

EFPDF

LGROTSG

• PHWR

r*SG

All-volatile treatment of secondary waterBrackish condenser cooling waterBoiling water reactorCanada deuterium uranium reactorCondensate demineralizationEffective FulJ jower daysFresh condensf " cooling waterLight-water-cooled graphite reactorOnce-through steam generatorsPressurized heavy water reactorPhosphate treatment of secondary waterSea-water-cooled condensersSteam generator

ISSN 0067 - 0367

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ISSN 0067 - 0367

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ATOMIC ENERGY FSLWM L'ENERGIE ATOMIQUE OF CANADA … · atomic energy of canada limited "team...

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Transcript of ATOMIC ENERGY FSLWM L'ENERGIE ATOMIQUE OF CANADA … · atomic energy of canada limited "team...