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.
13. Nuclear Power Experience, Vol. PWR-2, section V.D., p. 102, paragraph 229.
- 34 -
14. Nuclear Power Experience, Vol. PWR-2, section V.D., p. 119, paragraph 269.
15. Nuclear Power Experience, Vol. PWR-2, section V.D., p. 96, paragraph 220.
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
CondenserTubeMaterial
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
CondenserTubeMaterial
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 Ti
condenser beingretubed with CuNi
horizontal SG
condenserretubed fromAl Brass
condenser beingretubed with Ti
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 GENER^UOR 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
AtiJmuner^oexportBalcki'fl̂ hf-r.i-V K. Wi 1 ,-ovHabrock & Wilenx CanadaCorabnsrinn EngineeringCockiirll l[leutsche Bab.-nck i, WilcoxFr;imatomoFoster Wh^elt-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
SecondaryChemistryControl
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
CumulatedTube Defects
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
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
SecondaryChemistryControl
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
CumulatedTube Defects
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
boric acid added tosecondary water
CANDU, horizontal SG
CD not full-time,2 SG's tubed withstainless steel, LGR
OTSG
OTSG
OTSG
CD under construction
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
CumulatedTube Defects
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
SecondaryChemistryControl
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
CD under construction
CD under construction
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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Chalk River, Ontario, Canada
KOJ 1J0
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