METALLURGICAL PROPERTIES OF COLD-WORKED ZIRCALOY-2 PRESSURE TUBES IRRADIATED FOR FIVE YEARS IN THE NPD REACTOR - Atomic Energy of Canada Limited ...
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Atomic Energy of Canada Limited
METALLURGICAL PROPERTIES OF
COLD-WORKED ZIRCALOY-2 PRESSURE TUBES
IRRADIATED FOR FIVE YEARS IN THE NPD REACTOR
by
W.J. LANGFORD
Chalk River, Ontario
January 1970
AECL-3516METALLURGICAL PROPERTIES OF COLD-WORKED ZIRCALOY-2 PRESSURE
TUBES IRRADIATED FOR FIVE YEARS IN THE NPD REACTOR
by
W.J. Langford
A B S T R A C T
Zircaloy-2 pressure tubes were irradiated to a
maximum neutron dose of 1.2 x 1 0 2 1 n/cm2 (>1 MeV) at 10,500
psi hoop stress in the Nuclear Power Demonstration Reactor
(NPD) . Post-irradiation examination showed that mechanical
strength increased and ductility decreased as a result of
the reactor service. The tubes can sustain a 4 in. axia^u-
slit at operating stress without catastrophic propagation
of the slit, ensuring that, coolant leakage would precede
catastrophic fracture. Maximum deuterium uptake was 9 ppni
per annum. The rolled joints at the tube ends were sound,
and showed no evidence of leakage or corrosion.
All measured properties were consistent with require-
ments for a pressure tube life of 30 years.
Metallurgical Engineering Branch
Chalk River Nuclear Laboratories
Chalk River, Ontario
January 1970
AHC1.-3SH)Propriétés métallurgiques des tubes de force en zircaloy-2 écroui,
après cinq années d'irradiation dans le réacteur NPD
par W.J. Langford
Résumé
Des tubes de force en zircaloy-2 ont été irradiés
jusqu'à une dose neutronique de 1.2 x 10 2 1 n/cm2 (>1 MeV)
alors que l'effort de frette était de 10 500 lb/in2 dans le
réacteur NPD. Un examen post-irradiation a montré que la
résistance mécanique augmentait et que la ductilité diminuait
par suite du service en réacteur. Les tubes peuvent supporter
une fente axiale de 4 in. dans les conditions opératoires sans
propagation catastrophique de la fente qui provoquerait une
fuite de caloporteur puis une grave rupture. L'apport
maximal du deuterium a été de 9 ppm par an. Les joints
roulés aux extrémités des tubes se sont avérés en bon état
et n'ont montré aucune trace de fuite ou de corrosion.
Toutes les propriétés mesurées étaient conformes
aux exigences voulant que les tubes de force servent pendant
30 ans.
L'Energie Atomique du Canada, Limitée
Chalk River, Ontario
Janvier 1970
AECL-3516METALLURGICAL PROPERTIES OF COLD-WORKED ZIRCALOY-2 PRESSURE
TUBES IRRADIATED FOR FIVE YEARS IN THE NPD REACTOR
\
by
W.J. Langford
1. INTRODUCTION
The Nuclear Power Demonstration (NPD) reactor '
is fuelled by natural uranium and uses heavy water as cool-
ant and moderator. The core comprises 132 horizontal z i r -
caloy-2 pressure tubes spaced on a square l a t t i c e in a double-
walled aluminum calandria vessel. The fuel channels l i e in
the east-west direction, and coolant flow is reversed in
adjacent channels. Stainless steel end-fittings at both ends
of each pressure tube contain removable closure plugs for
on-load refuelling. Fig. 1 shows the general arrangement of
the fuel channels.
NPD has produced electricity since June 1962. In
April 1967 two pressure tubes with their end-fittings were
removed to enable zr-2.5 wt% Nb pressure tubes to be irradiated,
This report describes the post-irradiation examination of the
cold-worked zircaloy-2 pressure tubes, including c r i t i c a l crack
length measurements, tensile t e s t s , deuterium analyses and
biaxial burst t e s t s .
2. MATERIALS
The Z i r c a l o y - 2 p r e s s u r e tubes were extruded and c o l d
drawn t o give approximately 17% r e s i d u a l cold-work, followed
by a u t o c l a v i n g f o r 72 hours a t 400°C. Nominal tube dimensions- 2-
were 3.25 i n . inside diameter with 0.170 i n . wall. All tubes
were made by Chase Brass and Copper Company Limited, from
ingots supplied by Carborundum Metals Company. Ingot analyses
are given in the Appendix. The manufacturer's s e r i a l numbers
for the tubes removed from NPD were 10 and 13 0, from l a t t i c e
positions G-5 and H-6 respectively.
Each end of the pressure tube was mechanically expanded
by rolling into the hub of a stainless steel end-fitting, form-
ing a leak-tight j o i n t : no welds were used (see Fig. 1).
The figure also shows the aluminum calandria tube which sur-
rounds the pressure tube; an Inconel 750 helical garter spring
(not shown) acts as a central spacer, preventing contact between
the hot pressure tube and the cool calandria tube»
3. OPERATING CONDITIONS
The p r e s s u r e tubes were i n t e r n a l l y p r e s s u r i z e d by
heavy water c o o l a n t a t 252-273°C and 1100 p s i g . At t h i s p r e s -
s u r e t h e c i r c u m f e r e n t i a l (hoop) s t r e s s i n t h e p r e s s u r e tube
was 10,500 p s i , c a l c u l a t e d from
a. = ~ a, = hoop stress, psi
h 2t h
P = internal pressure, psig
d = internal diameter, in.
t = wall thickness, i n .
The annular space between the pressure tubes and their c a l -
andria tubes was f i l l e d with nominally dry a i r , but there is
evidence that water leaked into the annuli at some time.
The pressure tubes, garter springs and end-fittings
were removed from the reactor in April 19G7 after 5 years'
service, equivalent to 3-1/3 years of continuous full-power
operation. The maximum neutron dose received by the pres-
sure tubes was 1.2 x 10 2 1 n/cm2 (>1 MeV)(3).
4. OBJECTIVES AND TEST PROGRAM
Each p r e s s u r e tube was removed from t h e r e a c t o r i n- 3 -
six 28 in. long pieces, leaving a 6 in. stub of pressure
tube projecting from the east end-fitting (coolant outlet
end for Channels G-5 and H-6) while the tube was cut almost
flush with the face of the west end-fitting. Fig. 2 shows
schematically the two pressure L«"_-L and the location of tost
specimens used in the investigation.
4.1 Visual Examination
The pressure tube sections were carefully examined
for signs of damage or corrosion. Particular attention was
paid to the areas where the garter spring spacers had been.
in contact with the pressure tubes, and to the extreme ends
of the tubes from the expanded joints in the end-fittings.
4.2 Deuterium and Hydrogen Pickup
Deuterium is released during the oxidation of zir-
conium in 280°C heavy water, and is partly absorbed into the
metal. When the deuterium concentration exceeds the solid
solubility level («0.04 parts per million (ppm) by weight,
at 20°c(4)), the excess precipitates as platelets of zirconium
deuteride. Larga quantities of these precipitates adversely
affect the mechanical strength and ductility of zirconium
alloys, and so coolant chemistry is carefully controlled to
minimize deuterium pickup.
Hydrogen i s present (- 4 -
of this 'sufficiently large 1 defect is termed the c r i t i c a l
crack length of the tube.
•Hie ability of the NPD tubes to sustain damage was
studied using a s l i t - b u r s t test technique(5). A through-wall
axial s l i t (approximately 0.006 i n . wide) of known length
is cut into an 11 in. or 18 in. tube specimen by electrical
discharge machining. The s l i t closely approximates a real
crack, which is the most severe type of defect which could
be encountered. The s l i t is mechanically sealed, and the
assembly internally pressurized with nitrogen to failure.
Failure stress is calculated from Pd/2± where P = internal
pressure (psig), d = internal diameter (in.) and t = wall
thickness ( i n . ) . A series of tests is performed with dif-
ferent s l i t lengths, and the results plotted on logarithmic
scales as hoop stress at failure versus s l i t length. The
curve has the form a an = C, where a = hoop stress at f a i l -
ure (psi), a = s l i t length ( i n . ) , n and C are constants.
Extrapolation of the curve to reactor operating stress indicates
the c r i t i c a l crack length.
Three 18 i n . and six 11 i n . long irradiated NPD pres-
sure tube specimens were tested in this manner, together with
seven 11 in. specimens of unirradiated tube for comparison.
4.4 Biaxial Burst strength
When a tube with closed ends is internally pres-
surized, a biaxial stress system results in which the c i r -
cumferential (hoop) stress is twice the axial s t r e s s . The
ultimate hoop strength of such a tube, pressurized to failure,
is about 15% higher in isotropic material than that measured
in a uniaxial tensile test, due to deformation restraints
imposed by the stress system(^). Cold drawn Zircaloy-2 pres-
sure tubing shows about 20% increase in strength in biaxial
tests as a result of the anisotropy of deformation in the
highly-textured material(7).
Biaxial burst tests therefore give a more r e a l i s t i c
indication of reactor pressure tube properties than uniaxial
tensile t e s t s . Specimens from the irradiated NPD tubes were
tested in this manner and compared with unirradiated data.- 5 -
4.5 Uniaxial Tensile Properties
The uniaxial tensile t e s t is the conventional method
for evaluating strength and ductility. Irradiation of pre-
machined tensile test specimens provides a ready means of"
studying irradiation effects on these properties, although
i t does not easily permit study of the effect of stress during
irradiation. Longitudinal and circumferential (ring) test
specimens were therefore machined from irradiated NPD pres-
sure tubes and results compared with pre-irradiation data and
with results from experimental irradiations.
5. RESULTS AND DISCUSSION
5 .1 The Tube Surfaces
5•1•1 inside surface:
Before tube removal the inside surface of each pres-
sure tube was examined with a borescope, and surface rough-
ness measured. This inspection revealed fine axial scratches
about 0.001 i n . deep, scored by the fuel bundles when moved
during fuelling. No fretting or other significant damage was
seen. visual examination of the tube sections after removal
confirmed these observations.
(
5.1.2 Outside surface:
Outer surfaces were dull grey, with longitudinal
grey-white streaky deposits along the bottom (Fig. 3), possibly
due to accumulation of water in the gas annulus during the
early part of the reactor's l i f e .
The outside of tube 10 was marked with fine parallel
lines in a yellow-white deposit at i t s mid-reactor position,
extending through perhaps 45° each side of bottom dead centre
(Fig. 4), i . e . , where the pressure tube was in contact with
i t s garter spring. The lines extended axially for some
distance to one side of the main deposit showing how the tube
section was drawn through the garter spring during removal
from the reactor. The yellow discolouration of the deposit
is probably due to iron oxide deposited from out-reactor
piping. Accumulated rust deposits were seen on the stainless- 6 -
steel end-fittings.
Fig. 5 i s a close-up view of the mark on tube 10 prior
to sectioning. To f a c i l i t a t e oxide thickness measurements,
Zircaloy-2 metallographic samples are pickled in 6N HCl to
remove crud while leaving the oxide film i n t a c t . Examina-
tion of a pickled transverse section through the mark showed
that i t was confined to the crud layer: there was no measurable
wear (within ±0.0005 in.) of the oxide layer or underlying
metal (Fig. 6). No hydride precipitates were seen in cross-
sections of the tube wall near the mark. Specimens taken
from a location 180° away from the mark ( i . e . , from the upper-
most part of the tube) were identical in appearance.
The most reasonable interpretation of the mark is
that the garter spring impeded movement of the water in the
annulus, "so that the yellow-white deposit (Fig. 4) b u i l t up
around the garter spring, forming an impression where the
spring touched the pressure tube. A similar mark on tube
130 was much less pronounced than that on tube 10 and was
not examined metallographically.
5.2 The Rolled Joints
The pressure tube end in each rolled joint was removed
by axial.ly s l i t t i n g the surrounding stainless s t e e l hub into
two pieces. The pressure tube and hub surfaces which had
been in contact showed no evidence of coolant leakage through
the rolled joint, or of corrosion (Fig. 7).
The oxide thickness on the inside tube surface varied
between 2 and 4 microns (Fig. 8) of which the original auto-
claved film accounts for about 1 micron. No deuteride pre-
cipitates were observed (Fig. 9 ) .
Circumferential thin foils were seen between the hubs
of the inlet end-fittings and the pressure tube ends (Fig.
10). These ' f o i l s ' were examined after sectioning the rolled
joints: they were roughly 1/8 i n . wide and 0.005 in. to
0.010 i n . thick. The material was metallographically iden-
tified as Zircaloy-2, and thus had been part of the pressure
tube. Considerable twinning indicated that the foil had
been heavily deformed (Fig. 11) . No zirconium deuteride
precipitation was seen.- 7 -
Slight axial scoring of the tube surface, together
with the twinning in the zircaloy-2 foils, suggests that
the foils resulted from limited axial shearing of the pres-
sure tube surface against the stainless steel hub during
joint rolling. 'The presence of the foils did not affect
the leak-tightness or strength of the rolled joints. The
pullout strength of the irradiated rolled joints was 105,600
lbs compared with 82,000 lbs before irradiation( 8 ).
5.3 Hydrogen and Deuterium Analysis
The results of analyses using specimens from several
locations are given in Table 1.
Hydrogen concentration is within the specification
limits for reactor grade Zircaloy-2 pressure tubes. Deuterium
analyses indicate that maximum pickup occurred in the tube
ends inside the i n l e t end-fittings.
Table 1
Hydrogen and Deuterium Analyses:
(Figures are Average of 3 or More Samples)
Pressure Tube Location H D2
2
ppm
ppm
10 Centreline Flux 17 20
10 Garter spring Mark 19 15
10 180° from Mark 19 15
lu Outlet End, =s6 in. 9
from Rolled Joint /
10 Tube End from inlet End- 23
Fitting /
130 Tube End from Inlet End- 31
Fitting /
130 Centreline Flux / 11
130 Outlet End, PS6 in. from / 11
Rolled Joint
/ not analyzed- 8 -
The rate of deuterium pickup varies with the rat^
of oxidation. Initial pickup rate falls rapidly as the
oxide thickens. After transition in oxidation kinetics at
abovt 35 mg/dm2 («2.5 \m of oxide), deuterium is picked up
at a rate which is roughly linear with time. Since the
oxides measured (Pig. 8) are close to the thickness at tran-
sition, it is not possible to say with certainty that the
deuterium pickup rate of the first five years1 operation
will continue. Nevertheless, linear extrapolation at this
rate provides an upper limit to the probable deuterium pick-
up in the pressure tube after a 30 year life.
The maximum pickup of 31 ppm in 3-1/3 full-power
years extrapolates to 278 ppm deuterium after 30 years.
This is equivalent* to a hydrogen pickup of 139 ppm which,
with the initial hydrogen concentration of the tube material,
predicts a maximum 'effective hydrogen1 concentration of 139
+ 25 = 164 ppm after 30 years. The hydrogen analyses made
on NPD tubing provide no evidence that hydrogen has been
picked up from reactions with light water in the gas annulus,
and hydrogen pickup is therefore excluded from the extra-
polation .
5.4 Critical Crack Length Measurement
Slit burst tests were conducted at 20 and 300 C C.
Results, given in Table 2 and Fig. 12, include those from
unirradiated NPD tubing for comparison. The curves in pig.
12 show the relationship between slit length and failure
stress, and predict a critical crack length of more than 5 in,
in unirradiated tubing at NPD operating stress.
The zirconium-hydrogen system is normally described in
terms of parts per million (ppm) of hydrogen, by weight.
Since the deuterium atom is .twice the weight of the
hydrogen atom, measurements of deuterium concentration
should be halved if they are to be related to the zr-H
phase system. Thus 'effective hydrogen1 concentration
of material from a heavy water system is found by adding
half the deuterium concentration to the hydrogen con-
centration.- 9 -
Table 2
S l i t Burst Test Results:
NPD Zircaloy-2 Pressure Tubing
Tube n/cm2 Specimen Test Slit Failure
No >1 MeV Length Temp Length Stress
(in.) (°C) (in.) (kpsi)
10 1 .2 X io 21 18 300 3 17.6
21
10 1 .2 X 1 0 IB 20 2 28.6
10 8 X 1 0 20 18 20 4 16.3
130 1 .2 X 1021 11 20 3 17.2
130 1 .0 X 1021 11 300 2. 5 24.0
130 8 X 1020 11 20 2 5 21.3
130 8 X 1020 11 300 2 34.2
31F 0 11 300 2 24.8
31F 0 11 20 2 34.4
94B 0 11 300 3 19.0
94B 0 11 20 3 20.4
22F 0 11 300 2 .5 21.1
22F 0 11 20 2 .5 23.0
29F 0 1.1 3 00 3 18.7
5.4.1 Effect of Neutron Irradiation on Critical Crack
Length :
Five years 1 in NPD (1.2 x 10 21 n/cm2 >1 MeV) reduced
c r i t i c a l crack length slightly at 20 and 300°C, from over
5 in. to about 4 i n . at both temperatures. The s l i t length/
failure stress relationship is close to that established
using virtually identical Zircaloy-2 pressure tubing made
for the Douglas Point Reactor(S). The present investigation- 10 -
supports the evidence in reference (5) that irradiation under
stress causes a slight reduction in failure stress for a
given slit length at 20°C, while irradiation of unstressed
tubes led to higher failure stresses.
(9)
Aungst and Defferding found little effect of
neutron irradiation on the room temperature failure stress
of flawed KER Zircaloy-2 pressure tubing, but did not report
neutron dose. In both KER and NPD tubing, pre-irradiation
and post-irradiation critical crack lengths were very large
in relation to tube wall thickness.
5.4.2 Effect of Deuterium on Critical Crack Length:
The expected deuterium pickup of 278 ppm (=139 ppm F U )
in the reactor's lifetime will not reduce the critical crack
length of NPD pressure tubes to a dangerous extent. Irradiated
Douglas Point type pressure tubing containing 200 ppm hydrogen
had a critical crack length of about 4 in. at 20 and 300°C, at
10,500 psi operating stress( 5 ).
5.4.3 Significance of Critical Crack Length:
For safe reactor operation, the critical crack length
should be large enough that a defect will reveal its presence
by coolant leakage before reaching critical length.
A crack growing from a small defect on the surface
of a pressure vessel will have reached a length of approximately
twice the wall thickness when it has penetrated completely
through the wall. if this crack is stable under the principal
applied tensile stress, the pressurizing fluid will leak and
the vessel will remain intact. This is the "leak-before-break"
concept(10,11) which is now widely used in determining the
safety of many types of pressure vessels.
Applied to the NPD pressure tubes, this means that
for safe reactor operation a crack of length equal to twice
the wall thickness, or approximately 0.35 in., should be
stable at reactor operating stress. Thus the measured critical
crack length of over 4 in. provides a wide margin of safety.
A short defect arising in an NPD pressure tube, and growing
through the tube wall perhaps by fatigue or creep, would
leak heavy water coolant before becoming unstable. Leakage- 11 -
rates from experimental slits have not been measured, but
should be quite large from a crack approaching critical
size.
The NPD reactor is not equipped to detect leakage
from individual channels but the moisture content of the
reactor vault atmosphere is controlled, and dryers remove
excess moisture. A substantial rise in vault moisture result-
ing from pressure tube leakage should be indicated by increased
load on the dryers.
5.5 Biaxial Burst Strength
20
An 18 i n . section of tube 130, irradiated to 9 x 10
n./cm (>1 MeV) , was burst at 300°C. Results, together with
unirradiated 280°C data for comparison(12) a r e given in
Table 3.
Table 3
Burst Test Results:
NPD Zircaloy-2 Pressure Tubing
Tube Test 0.2% Yield Burst Radial Strain Circumferential
Temp Stress Stress* to Failure % Strain L O
°C kpsi kpsi Failure %
130 300 79 80 12.5 / 8
Unirrad. 280 56 61 30 >20
(Ref. 12)
* Burst Stress - Pressure x Radius/wall Thickness.
/ Or Higher. Fracture Surface Damaged During Bursting.
Compared with unirradiated material, irradiation to
9.x 10 2 0 n/cm^ (>1 MeV) increased circumferential 0.2% yield
and ultimate s t r e s s e s of NPD pressure tubing at 300°C by
41% and 36% respectively, with an attendant drop in d u c t i l i t y
These changes are similar to those observed in a cold-worked
Zircaloy-2 pressure tube' irradiated in a CRNL loop^ 13 ^. The- 12 -
burst strength and ductility of the NPD tube is almost
identical to that of the loop tube after equal fast neutron
doses. strength is about 8% higher than that reported for
cold-worked Zircaloy-2 tubes from PRTRH4)# which received
maximum neutron doses of only RilO^O n/cm2 (>1 MeV) .
Unirradiated cold drawn zircaloy-2 tubes show unusually
high circumferential and radial strains when biaxially stressed
at 300°C. This is explicable in terms of the anisotropy of
deformation resulting from preferred crystallographic orien-
tations developed during fabrication(15)m isotropic materials
show much less strain in biaxial tests since the stress system
restricts normal deformation.
During irradiation to 9 x 1 0 2 0 n/cm (>1 MeV) the
ductility of the NPD pressure tubing fell, and the irradiated
tube behaves more like an isotropic material. Yield stress
is much closer to ultimate burst stress after irradiation
(Table 3) severely restricting uniform ductility. Consequently,
yield at the thinnest wall section immediately precedes attain-
ment of the ultimate stress in that section, after which local
plastic instability leads to rupture. When the thinnest sec-
tion has reached ultimate stress, much of the tube has not
yet yielded, and general circumferential plastic deformation
is much less than in unirradiated tubing. Reference (13)
suggests that the plastic instability can be attributed to
mobile dislocations sweeping channels free of defects. The
moving dislocations absorb defects, reducing the stress needed
to move succeeding dislocations. The situation is unstable
since the stress at initiation of plastic flow exceeds the
stress required to maintain flow. The reduced ductility
after irradiation therefore signifies changed deformation
modes rather than embrittlement.
5.6 Uniaxial Tensile Properties
0.25 in. wide rings were machined from the ends and
centre of both pressure tubes, and longitudinal flat tensile
specimens from a high-dose section of tube 130. Fig. 13
shows the specimen details. The specimens were tested at
20 and 280°C, and Table 4 gives the results.
There is little difference in the circumferential
tensile properties of the two NPD pressure tubes. Only oneTable 4
Tensile Results: Irradiated NPD Pressure Tubes
Tube Specimen Neutron Dose Direction Test Temp 0.2% Y . S . UTS Reduction i n Elongation
Location / n/cm2 \ °C kpsi kpsi Area, % in 1", %
\>1 MeV/
10 Coolant 5.5 x 10 2 0 Circumferential 20 * 102.0 38 *
•I
10 Outlet End 11
280 58.5 63
10 Coolant 5.5 x 1 0 2 0 Circumferential 20 107.0 35
10 I n l e t End ii
280 - 63.3 57 -
10 sa Flux 1.2 x 1 0 2 1 Circumferential 20 - 108.3 35
10 Centre II II
280 66.1 50 -
_
130 Coolant 5.5 x 10 2 0 Circumferential 20 99.8 38
130 Outlet End 11 II
280 - 61.3 51 I
130 Coolant 5.5 x 10 2 0 Circumferential 20 107.4 40 -
130 I n l e t End M M
280 - 64.7 53
130 e= FlUX 1.2 x 1 0 2 1 Circumferential 20 104.5 36
130 Centre II if
260 - 69.5 54 -
130 as FlUX 1.2 x 10 21 Longitudinal 20 100.1 117.0 23 5.3
Ii
130 Centre 280 79.9 85.2 31 5.7
* 0.2% Y.S. and Elongation cannot be measured on ring tensile specimens.- 14 -
circumferential specimen was tested from each location at
each temperature: longitudinal results are the average of
three tests.
5.6.1 Effect of Temperature:
UTS is higher at the coolant inlet end than at the
outlet end, particularly in tests at 20°C. Neutron dose is
the same at each end, but outlet coolant temperature is 25°C
higher than inlet temperature(2). The back end of an extruded
and cold drawn tube is usually slightly stronger than the
front end(16), probably because the back end of the extrusion
billet is cooler at the moment of extrusion, leading to more
residual co\d-work in the back end of the tube. In NPD,
the back end of each pressure tube is at the east end of the
calandria, which for channels G-5 and H-6 is the coolant out-
let end. Assuming then that these ends were slightly stronger
at the time of installation, the difference in circumferential
properties between inlet and outlet ends of the irradiated
tubes is probably significant. The explanation may lie in
differential recovery of irradiation damage and ccld-work
due to the temperature gradient along the pressure tube.
5.6.2 Effect of Neutron Dose:
Tensile t e s t s showed that tube strength at 280°C
increased with fast neutron dose. Results at 20°C were in-
conclusive, tube 130 showing a slight decrease in strength
between the coolant inlet end and flux centre.
Based on Veeder's and Thomas unirradiated
1
NPD tube data, 5 years service in NPD has increased longi-
tudinal and circumferential tensile strengths while reducing
ductility, as shown in Table 5.
The irradiation-induced increase in circumferential
UTS agrees well with that of the zircaloy-2 pressure tube
irradiated to 1.2 x 10 21 n/cm2 (>1 MeV) at ss270°C in a
CRNL loop( 1 3 ). Reactor service reduced only slightly the
reduction in area measured in circumferential tension.
Longitudinal strength increased much more than c i r -
cumferential strength, the increase at 280°C being par-
ticularly large. A tensile t e s t on a spare irradiatedTable 5
Changes in Uniaxial Tensile Properties of NPD Zircaloy-2
Pressure Tubes Resulting from Reactor Service
Direction Tes t Temp Property Unirradiated Irradiated % Change
°C (1.2 x 1 0 2 1 n/cm2 >1 MeV)
0.2% Yield Stress, kpsi 71.0 100.1 +40
UTS, kpsi 89.0 117.0 +31
Longitudinal 20
Total Elongation, % 20.5 5.3 *
Reduction in Area, % 33 23 *
0.2% Yield Stress, kpsi 45 79.9 +77
UTS, kpsi 53 85.2 +60
Longitudinal 280
Total Elongation, % 22.5 5.7 *
Reduction in Area, % 45 31 *
0.2% Yield Stress, kpsi 86.0 Not Available
UTS, kpsi 90.0 104.5 + 16
Circumferential 20
Total Elongation, % 8.2 Not Available *
Reduction in Area, % 45 36 *
0.2% Yield Stress, kpsi 47.0 Not Available
UTS, kpsi 49.8 69.5 +39
Circumferential 280 Not Available
Total Elongation, % 21 *
Reduction in Area, % 60 54 *
Ductility changes not compared because of differing specimen types used.- 16 -
specimen confirmed the high longitudinal strength: metal-
lographic examination of the tensile specimen revealed no
unusual structural features. The increase in longitudinal
strength agrees with that reported for similar 18% cold drawn
tubing irradiated to 1.3 x 1 0 2 1 n/cm2 (>1 MeV) in water at
280°c( 17 ).
The anisotropic irradiation strengthening shown by
the NPD tubing indicates that the principal mechanism of
hardening is impedance of slip. Cold drawing of NPD Zircaloy-2
pressure tubes produced a preferred crystallographic texture
in which virtually all basal poles lie in the circumferential
direction^ 5 ) , so that under axial tension the material deforms
primarily by [10Î0] slip. When loaded in circumferential
tension, a small proportion of the grains will slip, but
most grains are oriented for {1012} twinning, which is probably
triggered by stress concentrations at boundaries between
slipped grains and grains which will twin.
After irradiation slip requires much higher applied
stresses than before. Longitudinal specimens, deforming
almost completely by slip, therefore show a large increase
in yield strength. Circumferential specimens show some
increase, as the slip-oriented grains require higher stresses
to initiate the slip. Once slip begins, however, the stress
buildup against grains oriented for twinning is high enough
to start twinning.
Longitudinal reduction in area fell with irradiation.
Elongation values are not directly comparable because several
different specimen types (and thus gauge lengths) were used
in testing NPD material for earlier reference properties.
5.6.3 Effect of Deuterium:
The predicted maximum deuterium concentration (278
ppm after 30 years) will have little effect on tensile pro-
perties of NPD coolant tubes at operating temperatures.
Sawatzky(18) found little effect of 200 ppm hydrogen (3 400
ppm deuterium) on strength up to 400°C, although below 150°c
ductility was reduced. The combined effect of 200 ppm hydrogen
and neutron irradiation to 2.3 x 1 0 2 0 n/on2 (>1 MeV) on the
strength of Douglas Point tube material was virtually in-
distinguishable from the effect of irradiation alone at 20- 17 -
and 3OO.°C (5) .
6. CONCLUSIONS
(1) After 5 years of reactor operation the critical
crack length of the NPD Zircaloy-2 pressure tubes is over
4 i n . at operating s t r e s s . This is ample to ensure that a
through-wall crack would leak coolant before attaining
c r i t i c a l size.
(2) Biaxial burst strength at 280°C increased by 30%.
Circumferential tensile strength increased by 40%, and longi-
tudinal tensile strength by 60%, at 280°C. Ductility fell
but remains sufficient to permit considerable local plastic
deformation at a stress concentration, preventing b r i t t l e
(low stress) failure.
(3) Deuterium pickup is at most 9 ppm per year at full
power, indicating a maximum concentration of 278 ppm after
30 years' full power operation. This concentration should
not substantially affect the strength or defect tolerance
of the pressure tubes.
(4) Contact with the garter spring h^s not measurably
worn the pressure tube.
(5) The rolled joints show no signs of leakage or cor-
rosion. Thin foils surrounding the pressure tubes at the
i n l e t end-fittings may have been produced during the joint
rolling operation. The foils do not affect the safe operation
of the fuel channels.
7. ACKNOWLEDGEMENTS
The author is pleased to acknowledge the experimental
assistance of L.E.J. Mooder, J.G. Bryson, J.E. Kelm and
D.N. Peck.
8. REFERENCES t
(1) Nuclear Power Demonstration Reactor, AECL-1634,
October 1962.- 18 -
(2) Directory of Nuclear Reactors. IV, 151-156 (Vienna,
IAEA, 1962) . Also available as AECL-1330.
(3) P.C. Barnett f "Calculation of Fast Neutron Flux in
Reactor Pressure Tubes and Experimental Facilities",
AECL-3167, July 1968.
(4) J.J. Kearns, J. Nuc. Mat* June 1967, Vol. .22.. P- 292.
(5) A. Cowan and W.J. Langford, "Effect of Hydrogen
and Neutron Irradiation on the Failure of Flawed
Zircaloy-2 Pressure Tubes", J. Nuc. Mat. April 1969,
Vol. 3_£, No. 3, p. 271. Also available as AECL-3262.
(6) 0. Hoffman and G. Sachs, "Introduction to Theory
of Plasticity for Engineers", McGraw-Hill, 1953,
p. 39.
(7) P.A. Ross-Ross and K.L. Smith, "Pressure Tube Develop-
ment for Canada's Power Reactors", 64-WA/NE-5, ASME
Winter Annual Meeting, 1964.
(8) E.C. Carlick and K.W. Roche, CRNL-151, AECL Un-
published Report, June 1968.
(9) R.C. Aungst and L.J. Defferding, "Crack Propagation
Tests on Zircaloy-2 Reactor Pressure Tubing in Both
the Normal and Hydrided conditions", ASTM Special
Technical Publication 380, p. 451, June 1965.
(10) Fifth Report of the Special ASTM Committee on Frac-
ture Testing of High-Strength Materials, Materials
Research and Standards, _14_, No. 3, 1964.
•
(11) G.R. Irwin and A.M. Sullivan, Proc. Roy. Socw, 285A
X
(1965), p . 141.
(12) w.R. Thomas et a l , " I r r a d i a t i o n Experience with
Zircaloy-2", Third U.N. I n t . Conf. Peaceful Uses
Atomic Energy, Geneva, September 1964 (AECL-2021).
(13) W.J. Langford, "Metallurgical Examination of a
Cold-Worked Zircaloy-2 Pressure Tube I r r a d i a t e d
in the U-2 Loop", AECL-3457, October 1969.- 19 -
(14) M.C. Fraser, " P o s t - I r r a d i a t i o n Evaluation of
Zircaloy-2 PRTR Pressure Tubes - Part I I I " , BNWL-5,
January 1965.
(15) B.A. Cheadle and C.E. E l l s , Trans. Met. Soc. AIME •
233., 1965, p . 1044.
(16) J . Veeder, "Tensile P r o p e r t i e s of NPD-2 Pressure
Tubes", AECL-1247, April 1961.
(17) J . E . i r v i n , Eleccrochem. Tech. £, 5-6, May-June
1966, p. 240.
(18) A. Sawatzky, "Effect of Neutron I r r a d i a t i o n on the
Mechanical Properties of Hydrided Zirconium Alloys",
AECL-1986, June 1964.- 20 -
APPENDIX
INGOT ANALYSES FOR THE NPD PRESSURE TUBE MATERIALS
Analysis, ppm by weight
Element Tube 10 (Channel G-5) Tube 130 (Channel H-6)
ingot K 223 ingot K 304
Al 28 31
B 0.5HEAVY WATER MOOERATOR LEVEL
HELIUM GAS
BALANCE LINE
LIGHT WATER REFLECTOR
CALANOBIA :
ENLARGED VIEW OF ROLLED JOINT
-OUTER END WALL
- INNER END WALL
-SPRAY SVSTEM
Zircaloy 2 / COOLANT TUBE ASSEMBLY
pressure tube
NPD REACTOR
ARRANGEMENT
FLOW O U U E '
f N O - FITTING
(STEEL)
FU€L CHARGE
CLOSOBt PlUG »SS£MB(.» DlSCM*»GE
CLOSUBf P.UG 4SSEMBL»
PRESSURE TUBE ASSEMBLY
FIG.1RT C C C C LT R T O RTO
TUBE 130
(CHANNEL H-6)
WEST COOLANT FLOW EAST
to
6 5 4 3 2 1
1 I. 1I I TUBE 10
(CHANNEL G-5)
1 1 1 1 i
0 HT ORT C M 0 C C RT 0
FIG.2 LOCATION OF TEST SPECIMENS;
KEY: 0 OXIOE THICKNESS
RT RING TENSILE NPD Z I K C A L O Y - 2 PRESSURE TUBES
C CRITICAL CRACK LENGTH
LT LONGITUDINAL TENSILE
B BIAXIAL BURST
M METALLOGRAPHY (GARTER SPRING MARK)
1to6 ORIGINAL SECTIONS- 23 -
FIGURE 3 . Grey-White Surface Discolouration on
Bottom Outer Surface of NPD Zircaloy-2
Pressure Tube- 24 - P i g . 4 . Garter Spring Mark: NPD z i r c a l o y - 2 Pressure Tube No. 10 (Channel G-5) . Xl Approx. Pig. 5. Close-up View of Garter Spring Mark, X2.5 Approx.
- 25 -
(b)
FIGURE 6 . Circumferential Cross-Section of NPD Zircaloy-2
PressureTube No. 10 at Garter Spring Mark
(a) x 7.5 (b) x 50- 26 -
FIGURE 7 . Surface of Zircaloy-2 Pressure Tube Stub
Removed from NPD Inlet End-Fitting No. 136
(Channel H-6). Arrow Shows outboard End of
Tube.- 27 -
FIGURE 8 . Oxide on I n s i d e Surface of NPD Z i r c a l o y - 2
P r e s s u r e Tube No. 10 (Channel G-5), 2-4 ^m thick.
x 1000 V- 28 -
Radial
Direction
Vf
-S. . .
Longitudinal
Direction
Fig. 9. Microsection of NPD Zircaloy-2 Pres-
sure Tube No. 10, Taken from the Rolled Joint
Stub. Dark Stringers are Zr-Sn Intermetallic
Compound: There are No Precipitated Hydrides.
Bright Field Illumination. X200.- 29 -
End-Fitting
Pressure Tube Stub
FIGURE 10. Zircaloy-2 Foil (Arrowed) at Pressure Tube/End-
F i t t i n g Interface NPD i n l e t End-Fitting No. 136
(Channel H-6).- 30 "
FIGURE 11 . Zircaloy-2 Foil From Pressure Tube/End-Fitting
Interface, Channel G-5 Inlet End (Tube No. 10,
End-Fitting No. 304). x 250 Polarized Light- 31 -
50
300°C
40 j
30
WIO 2 ' n/tm1
25 UNIRRADIATED
20
!5
NPD
OPERATING
STRESS
£ io
se
tn
UJ
a.
en
20°C
w 60
Z)
5 50 A--- «S 10 n/tm
• UNIRRADIATED
40
30
25
20
NPD
OPERATING STRES
10 -L J l
.5 2 2.5 3 4 5
SLOT LENGTH INCHES
FIG. 12 FAILURE STRESS AS A FUNCTION OF SLOT
LENGTH NPD ZIRCALOY-2 PRESSURE
TUBING- 32 -
0.50rad -J- I— .no"
I— 0.25"
Fig. 13 TENSILE SPECIMENS USED IN THE INVESTIGATIONAdditional copies of this document
may be obtained from
Scientific Document Distribution Office
Atomic Energy of Canada Limited
Chalk River, Ontario, Canada
Price • $1.00 per copy
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