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Transcript of Nuclear Pressurized Water Reactor (PWR) - Mechannical design materials
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XA9949546
3. REACTOR PRESSURE VESS EL MATERIALS
K. Suzuki, The Japan St eel
Works,
Ltd., Muroran Plant
1 INTRODUCTION
The demands placed on rea ctor pressure vessel (RPV) steels are
severe. They must be manufactured in required sizes and
thicknesses, be of sufficie nt strength and toughnes s, show little
deterioration under irradiation, allow the production of high
quality welds and be compati ble with the cladding. This Chapter
refers to non-WER press ure vess els, but many of the guiding
principles described here apply to that particular case.
Starting with carbon steel plates and forgings for conventional
boiler drums in the dawn of commercial light water reactors (LWRs)
followed by a few change s thereafter, SA533 and SA508 and similar
grade steels have become well established [1 ]. Both grades are of
the vacuum treated, que nche d and tempered type of 600 N/mm
2
strength
class, which is not the highest level in weldable structural steels.
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The specific requirements for RPV steels are to giv e, even to l arge-
size component materials of
R P V s ,
higher values of the following
properties:
- uniformity and isotropy of mechanical properties, including
less mass effect in the mid-section
- fracture toughness
- internal defects
- weldability
- resistivity to neutron irradiation embrittlement
Some further items were added during the past two decades primar ily
for the purpose of easier execution of non-destructive examinations
both pre-service and in-service. These are :
- fewer weld seams in RPV
- larger and more integral design of component materials
The requirements have been steadily realized.
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With respect to the weld, improvements have been made by adequate
selection of flux in combination with requirements on the chemical
composition of the weld wir e, welding parameters such as weld bead
size and seque nce to minimize the coarse grained areas in the heat
affected zone (HAZ) [2 ].
2 HISTORICAL REVIEW [3]
Many of t he earlier plants were constructed of so-called carbon or
mild steels (usually in normalized and tempered, NT, c o n d i t i o n ) , but
several of these were prototypical units and applications have moved
to the wid ely accepted low alloy mangan ese, molybdenum, nickel,
quenched and tempered (QT) grades of higher strength. Early gas
cooled reac tors were also of carbon steel, but these were replaced
in later vers ions by prestressed concrete vessels.
Most of th e reacto rs operating at mid-197 3 were constructed of the
manganese-mo lybdenum steels in QT conditio n. In practice, the use
of the low alloy steels has predominated. For example, all of the
Japanese react ors o f BWR and PWR types have A533-B Class 1 (QT)
pressure ve ss els except Tsur uga which uses A302-B (QT) and JPDR-2
with A302-B in NT condition [4] . The sole Japanese gas cooled
reactor (Tokai Nuclear Power Sta tion) was contained in a vessel of a
Japanese C-Mn stee l. Typica l composition data for the Japan Steel
Works
( J S W ) ,
Ltd. steel use d is C-0.10*, Mn-1.30%, Si-0.25%, P-
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0.014%, and S-0.0 18* (equivalent to the UK steel, BS NDIV originally
planned for this reactor
v e s s e l ) .
Special care was taken to
minimi ze p hosp horus and sulphur and to refine grain size, thereby
enhancing no tch toughness. In addition, an experimental programme
was conduc ted to develop the best welding electrode material for
joining plat es of the Tokai vessel [ 5] . The relative content of
mangane se and silicon was determined for optimum toughness of welded
structures.
The Ag esta reactor of Sweden was contained in a carbon steel,
equivale nt to ASTM Type A212-B (a carbon-silicon-manganese
s t e e l ) ,
the Osk ars hamn- 1 vessel s teel was equivalent to ASTM Type A302-B [6]
(a manganese-mo lybdenum
s t e e l ) .
The A212-B steel was also used in
on e of t he e arl y U.S. reacto rs, Indian Point-1, and a limited number
of experimental reactors.
In the Fede ra l Republic o f Germany (F.R.G.) the steel designations
matc h nati ona l conventions, but most water reactor steels were quite
similar to U .S . grades A508-C1.1 for forging and A533-B for plate
steel [7 ]. And the most RPVs were strongly dependent on the 22
NiMoCr 37 composition (similar to A 5 0 8 , C1.2 : a nickel-chromium-
molybdenum
s t e e l ) .
22 NiMoCr 37 steel was used in F.R.G. until 1976
and fulf illed all requirements [8 ]. However, this type of steel
exhibited so me susceptibility to stress relief cracking and
underclad cracking [ 9 - 1 1 ] . By this reason the use of 20 MnMoNi 55
steel (similar to A 5 0 8 , C 1.3 : a manganese-molybdenum-nickel steel)
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was increased.
Little basic difference in steel type exists among the vesse ls of
reactors now operating throughout the world. Howev er, certain
metallurgical differences of crucial importance are identificable,
but other considerations such as neutron flux and fluence,
irradiation temperature, stress state etc. have equa l o r gre ater
bearing on radiation embrittlement sensitivity. Relati ve to
radiation embrittlement, however, it is important to realize t hat
a l l s t e el s u s e d t o d a te h a v e t h e s am e b a s ic w e a k n e s s . T h e y a r e
susceptible to radiation hardening, increases in stre ngth, and
reduction to fracture toughness. Thus, it is appropriate, eve n
necessary, to generalize initially in describing radi ation
embrittlement of vessel steels.
The curr ent choice of material is of specific A-53 3B, Cl.l allo ys
f o r p l a te s a n d A-508, C1.3 alloys for forgings (Table 1 ) [1 2] .
These ar e ve ry similar steels of the low alloy m anganese-mo lybdenum
type (A302-B) with nickel added. Significant differences have be en
developed between the old and new steels. Major differences ar e in
improved strength and toughness for the new s teels with a change of
micro -struc ture from basically pearlitic to tempered bainitic t o
temperecd martensiti c or mixed bainitic-martensitic microstr uctur e.
Chemically, the new steel is purer, that is, it contains smaller
amounts of r esidual or tramp elements, because steel-making
procedures incorporating vacuum degassing steps are applied
routinely in their production. These steps , producing singifican t
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changes,
are considered improvements from the vi ew o f r adiati on
embrittlem ent, but the resulting implication is that the gr eates t
concern must be addressed to the older plants , where our knowled ge
of the alloys used and the service conditions is least. Know ledge
about welds is probably both the most critical a nd the most limited.
Emphasis then must be placed on learning mor e about the old systems
material s, especially welds , and improving the new (future)
materials.
3 BASE MATERIAL
3.1 Chemi cal Composition [12]
Specifications for the structural steels whic h ar e extensiv ely used
for the LW R components in Germany and in the Unite d States ar e
listed in Table 1 [12].
Som e basic investigations were conducted to confir m th e effect of
chemical composition on 20 MnMoNI 55 in addition to the basic
studies for heavy forgings [10, 1 3 ] .
Fig.l s hows the continuous cooling transfor mation curves (CCT curve)
for this steel with 0.17% and 0.20* C. The differences in ferr ite,
pearlite and bainite transformation range ar e obser ved due to a
little difference in C content. Also the effect of M n, Ni, Mo on
the hardenability was investigated, but the res ults d o not indica te
substantial differences in the CCT curve.
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In order to main tai n sufficient toughness o f large forgings at lower
temperature , it is important to provide fine gr ain as well as proper
structure.
There a re some procedures to refine the grain size, and
it is an usual procedure to add Al. During the quenching of large
and mass ive for ging , austenitizing requir es considerably long time
to obtain uniform heating. In this case, the presence of AIH is
strongly effective in preventing grain growth.
V is know n as a predominant element to enhance the tensile
properties of material. However, the toughness and weld crack
sensitivity o f the material are also affected by the addition of
this elemen t. Thus no V should be added, even though the maximum
content of 0.03 or 0.05% is allowed in the mater ial specifications-
It is said th at Cu and P affect the irradiation damage which is
evaluated by the shift in Charpy impact curve [1 4] . In response to
this requir ement, the target values of Cu and P content of 0.0835
max.
and 0.0 08% max . respectively are maintained.
Concerning weldabi lity of the steel, liquation cracking and stress
relief cra cking in particular, crack promoting elements and
threshold value s for cracking occurrence were investigated.
Cracking ca n be avoided b y limiting the Mo cont ent, the elements P,
S, Cu, Sn, N, As , Co and Al as well [11] .
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3.2 Steel and Ingot Making [15]
3.2.1 Manufacturing Process of Large Ingots
(1) History o f the production of large ingots
In F i g . 2 , majo r changes in the history of the production of large
ingots in JSW are summarized. It also shows the transition of
maximum ingot size. Before Bochumer Verein-type vacuum casting
facilities wer e installed in 1959, steel was melted by acid open
hearth fu rnaces to minimize hydrogen pick up and the steel cast in
air.
After t he installation, acid open hearth was replaced by basic
open hearth and electric arc furnace ( E A F ) , because hydrogen removal
was made pos sibl e during vacuum casting. In order to obtain higher
degree of vac uum during casting, a steam ejector was introduced in
1 9 7 0 . In 197 3 a holding furnace was installed to replace the open
hearth. Accor ding to the demand from industries, sizes of ingot
became lar ger year after ye ar . In 1969 a world largest 400 ton
ingot was pr oduced, and the record was soon renewed by 500 ton ingot
in 1971, 570 ton ingot in 1980 , then 600 ton ingot in 1985.
(2) Installation of ladle refining furnace (LRF)
In 1980, to m ee t higher requirements for the record 570 ton ingots
of nuclear ve ss el application, vacuum facilities were installed to
the holding furnace to convert it into a ladle refining furnace.
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Fig.3 shows a schematic outline of the furnace. It has on e heatin g
system and two vacuum systems so that two vessels o f mo lten steel
can be treated at a time. Combining vacuum treatment at the LRF
with conventional tap degassing the "double degassing" pr ocess was
developed for the production of large ingots up to 600 ton. Fig .4
shows the production sequence of a 600 ton ingot. By this pro cess
the quality of products such as for nuclear applicatio ns wa s
remarkably improved.
(3) Gene ral description of large forging ingot
Fig. 5 shows a sulphur print of the longitudinal secti on of a 75 to n
ingot (left) and the description of segregations and solidif ication
structures o f the ingot
( r i g h t ) .
In a sulphur print solute-enriched
portions ar e marked dark. In the region called "branched column ar
zone", string-shaped A-segregates are observed and in the re gion
called "equiaxed
zone",
V-segregates are observed. In these two
regions, microporosities are easily formed. Non-metallic
inclusions , especially oxides, are often observed in the re gion
called "sedimentation
zone"
at the bottom of an ingot and thi s m ay
lead to the rejection of the entire ingot. Distribution o f C and of
other alloyi ng elements is not uniform. Generally speaking, thes e
elements a re rich in the top side of an ingot and poor in the
sedimentatio n zone. In the following section, technical
improvements to reduce these heterogeneities will be descr ibed.
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3.2.2 Rec ent Improvement in Steelmaking and Ing otmak ing
(1) Removal of P and S
Low S and P contents about 0.004-0.008% can be obta ined by
conventional EAF process, however, LRF process wa s introduced to
obtain extr emely low S and P contents in the liquid s t e e l . F i g . 6
shows a flow chart of the refining process by LRF. A high degr ee of
desulphurization up to 90 % can be obtained by using high bas ic slag
and extreme ly low content of S less than lOppm can be obtai ned. Th e
lowest recorde d S content is 2 ppm. Since P cannot be remove d by
LRF tre atment, it is essential to cut off oxidizing slag fro m EAF to
prevent rephosphorization. Fig.7 shows the changes of P content an d
S content whe n extremely low content of P+S
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(3) Degassing (H, 0) and reduction of non-metallic inclusions
(a) Effect of LRF
Degassing is achieved ver y effectively by LRF process due to high
vacuum and intensive stirring of molten steel. For the stirring, Ar
gas is blown into molten steel through porous plug at the bottom of
refining ladle. Energy dens ity of Ar stirring *., (Watt/ton) is
expressed by :
r
M
= 6
'
1 8 Q T
L (ln(l+
H
) + ( -
T
o ) }
148Pa T
L
Where
Q is flow rate of Ar-gas (Nm
3
/ m i n ) ,
Pa is pressure at the surface of molten steel ( a t m ) ,
T^ is temperature of mol ten steel (K) ,
T is temperature of Ar gas before blowing (K ),
M, is weight of molten steel ( t o n ) ,
H is bath depth (cm) ,
and t is treatment time
( s e c ) .
Fig.8 shows th e effect of
g
M
»t
on the degree of hydrogen removal.
In this process a high deg ree of hydrogen removal, up to 80% , is
obtained by increasing the "stirring" intensity e
M
»t , Fig.9 shows
a change in hydrogen content during LRF process . When molten steel
is cast by botto m pouring, about
0.5ppm
of hydrogen pick-up should
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be ex pected, whereas if it is cast by mold stream degassing, low
hydrogen levels of 0.4 to
0.6ppm
are obtained- Fig.10 shows the
e f f e c t o f
f
M
» t
o n t h e d e g r e e of d e o x i d at i o n . A t t h e r eg i o n o f l o w
£
M
«.t
degree of dexidation increases with increasing jw»t
however , it decreases with further increase of
£
M
»t »
It is
a t t r i bu t e d t o t h e e r o s i o n o f b r i ck o r s u s p e n s io n o f s l a g i n t h e
m e l t .
By controlling the stirring intensity g
M
«t a r o u nd 8 0 - 1 0 0 x l 0
3
J/ton, low oxygen level of less than 20ppm is obtained.
( b) E f f e c t o f do u b l e d e g a s s i n g
A s d e s c r i b e d b e f o r e , t h e d o u b l e d eg a s s i ng p r o c e s s ( va cu um t r e a t m e n t
a t L H F p l u s m o l d s t r e a m d e g a s s i ng ) i s v e r y e f f e c t i v e f o r d e g a s s i n g
o f m o l t e n s t e e l a n d w a s a p p li e d t o t h e p r o d u c t i o n o f t w o 5 7 0 t o n
i n g o t s o f n u c l e a r a p p l i c a t i o n s . F i g . 1 1 s h o w s t h e c h e ck a n a l ys e s o f
h y d r o g e n a n d o x y g e n o f t h es e i n g o t s . L o w h y d r o g e n c o n te n t s o f l e s s
t h a n l p p m a n d o x y g e n c o n t e n t s a r o u nd 1 0 t o 2 0 p p m a r e o b t ai n e d i n t h e
b o d y o f i n g o t s . F r o m t h i s l o w g a s c on t e nt t h e e f f e ct o f d o u b l e
d e g a s s i n g i s e v i d e n t .
( 4) R e d u c t i o n o f m a c r o s e g r e ga t e s
F o r 0 . 7 * c a r b o n s t e e l , c r i t i ca l c o n d it i o n f o r t h e f o r m at i o n o f A -
s e g r e g a te s i s e x p r e s s e d b y :
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Where
j is cooling rate in the radial direction of an ingot (°c/min)
and R is solidificatin r ate in the radial direction of an ingot
( m m / m i n ) .
By changing the constant term in the right side, the equation i s
applicable to other kinds of steel. If carbon content and silico n
content ar e hig h, the constant is large and A-segregates are easily-
formed. Mo and Cr have the opposite effect. In general it is
difficult t o control j and R during the solidification of an actual
ingot,
theref ore, chemical compositions are adjusted to minimize the
formation of A-segregates. Since the driving force for the
formation o f A-segregates is considered to be the density diff eren ce
between solute-enriched liquid and bulk liquid, the density
differnece ApL
i s
calculated assuming solute enriched liquid is
that at fra ction solid 0. 3* ApL varies with various steels and if
is large A-segregates are easily formed. The relation between
1 1 a n d A pL
i s
shown in
F i g . 1 2 .
If type of steel is deter mined ,
is calc ula ted and critical value j.R
1
- * is obtained from
Fig.12 and then the position where A-segregates first starts to fo rm
is estimated.
V-segr egates are formed due to rapid shrinking along the axis of an
ingot at the final stage of solidification. V-segregaties can be
reduced by ma king the taper of ingot larger and height-to-diameter
ratio H/D sm aller.
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(5) Redu ctio n of microporosity
Micropo rositi es are formed in the A-segregation zone and V -
segregation zone. If sizes of pores are large, it may result in the
rejection of the product. Therefore, it is very important to des ign
the optimum shape of the ingot to reduce porosity. Based up on
investigations on 3 - 220 ton ingots of carbon steel and low-all oy
steel,
critica l conditions for the formation of microporosity wer e
determined as shown in Fig.
1 3 .
Large porosity tends to form
especially in the V-segregation zone.
The length of the porosity zone Vy (mm) is expressed as a funct ion
of height-to-diameter ratio H/D as shown in Fig. 1 4 . In the figure,
total weigh t of ingot is kept constant and H/D is changed and it is
kno wn that Vy decreases with decreasing H/D. Porosity along t he
axi s of ing ot can be reduced also by increasing the weigh t o f th e ...
feeder hea d. In this case €he rate of axial solidification beco mes
slower and the formation of large porosity are suppressed. Bas ed
up on the abo ve mentioned studies, optimum ingot shapes were d esig ned
and satisfact ory results obtained.
(6) Reduc tion of carbon segregation
Quantita tive understanding of carbon segregation of ingots is ver y
important. However, studies to date have yielded improve ments,
because c ar bo n segregation is affected by many complicated fact ors .
The e quation given below shows the result of multiple regres sion
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analysis on 75 to 570 ton Mn-Mo-Ni steel ingots (number of ingot :
8 1 ) .
Segregation Rate [ - ( S Q - Q K ) adle
x 1 Q 0 ]
(*C)ladle
= 38.5+0.132(Wt)-72.8(F.H.R.)-100(/\C multiple)
Where
(%C) is check carbo n analysis i n t h e centre of (ingot
body/feeder head) boundary face
( w t & ) ,
(%C)
l a d l e i s
ladle analys is
o f
carbon content
( w t % ) ,
(Wt) is total ingot weight ( t o n ) ,
(F.H.R.)
is
feeder head ratio defined
a s
(weight
o f
feeder
head)
/
(weight
of
ingot body )
and (̂ C multiple) is carbon content differenc e at multi-pouring
practice defined as
£
{| (%C)i-(*C)
ladle |
x W i }
i - 1
i - 1
where W i i s t h e weight and (9SC)i is the carbon content o f molten
steel cast as "i"th time.
From the equation above, it is known that large feeder head ratio
and large carbo n content differences favour
t h e
minimizing
of
carbon
segregation
o f
ingots. Fig. 15 shows
t h e
distribution
of
carbon
content alon g the axis o f a 1 4 0 t o n a n d a 1 8 0 t o n rotor ingot cast
by multi-pouring method.
It is shown that these distributions a r e uniform compared with
ingots cast b y conventional method s. Fig.16 shows the distrubution
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of carbon content in the longitudinal section of 570 ton ingot for
nuclear application. The carbon segregation is not as severe as in
a "huge" ingot . It is considered that this results from the pouring
process.
If this ingot had been poured without a carbon content
difference, the carbon segr egation would have been about 109s
( 0 . 0 2 9 J C )
higher according to the estimation by the equation given
above.
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3.3 Forging Process [16]
F i g s . 17 and 18 show the combined vessel flange and nozzle belt
forging of KWU/1300MWe pressurized water reactor pressure vesse l
(PWRPV) made from a 400 t on ingot and the combined vessel flange and
nozzl e belt forging of WEC/1 57" PWRPV made from a 500 ton ingot
(developed by COCKERILL) as compared with the conventional one,
respective ly. These integrated flange forgings were hot worked by a
10,000 ton forging pre ss. Fig. 19 illustrates the forging proce sses
for the flange forging ma de from a 500 ton ingot. Sufficient
discard was made from each end of the ingot to insure that only
sound metal enters the completed forging. After piercing of ingot
c o r e , repeti tion of enlarg ing and upsetting was performed to close
possible porosity inside the ingot.
Th e for ging ra tio and repetition of enlarging and upsetting
operatio ns play an important role to improve the mechanical
prop erti es. Fig.20 shows the improved impact value due to
repetit ion of forging. Theref ore, from the results a minimum
forging r atio of 1.5 s hould b e required.
Anisotr opy o f forged materi al was intensively investigated by W.
Coupette in 1940th [17 ]. Fig.21 shows the effect of forging ratio
and anisotr opy of 20 MnMo Ni 55 steel. Compared with Coupette's
dat a, anis otro py of forging at present time seems to be much
smaller.
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To es timate the mechanical properties due to differences in forging
ratio, the use of the logatithmic strain concept is convenient. The
logarithmic strains in three directions are defined in
Fig.22.
From the results of mechanical tests for different components, the
relation betwe en logarithmic strain and mechanical pro perties, such
as tensile properties and Charpy V-notch impact value , can be
obtained, as shown in
Fig.23.
Tensile strength, yield strength and
elongation, at both room temperature and 350°c, are not so
significantly related to logarithmic strain, but the reduction of
area is increased with the increase of logarithmic strain.
Charpy V-notch impact values are also increased with the increase of
logarithmic stra in. The integrated flange forgings made from a 400
ton and a 50 0 to n ingot respectively were manufactured under the
above consideration. The logarithmic strains of integrated flange
forgings were as given below:
a) Shell flange :
£
t = Q . 9 6
£ a
= -0.28
fit - -0.68
b) Mono-blo ck vessel flange : £ t = n ni
ea = -0.18
fi
r
= -0.63
Therefo re, anisotropy in three directions is expected to be
minimized.
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3.4 Heat Treatment [16]
Heat treatment technology for large forgings is based on the
experimental investigations and experiences of actual ope ratio n in
addition to heat treatment theory. The important points to be
considered are t he segregation in large ingots, mas s effect s,
hydrogen induced defects, temper embrittlement and residual
stresses.
Flake,
ghos t crack and fish eye etc. are well known hydrogen induced
defects. Recentl y, the problems in large forgings due to hydr oge n
are remarkab ly decreased by the improvement of vaccum treat ment
technology.
However, it is still necessary to consider the prevention of
hydrogen induced defects depending on the size and grad e of
materials.
In general, the measures taken for the preven tion o f
hydrogen induced defects are as follows.
- Slow cooling after forging or rolling
- Isoth ermal annealing (Pearlite transf ormat ion)
- Normali zing and tempering (Bainite trans formation)
For large pie ces o f low alloy steel, normalizing and tem pering
technique is usuall y applied. Preliminary heat treatment after
forging sh own in Fig.24 is one example adopted in Germany for
preventing hydrogen flake [ 18] . In this procedure, the bainite
transfor mation is mainly performed after hot working of the fo rgi ng.
At pre sent, in JSH, the basic idea behind this heat tr eatment
diagram is als o being adopted for the preliminary heat tr eatme nt o f
large forgings such as integrated
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flange forging made from 20 MnMoNi 55 steel under the consider ation
of CCT curve shown in Fig.l.
After preliminary heat treatment (normalizing and t e m p e r i n g ) , the
forgings a re machined to a simple cross-section for perfo rming
ultrasonic examination. After successful examination, the forgings
are contour machined further for quenching and tempering in order to
obta in a go od quenching effect. The austenitizing temperatu re of
870 to 910 °c is selected to minimize the grain growth and a faster
quenching operation is for a more complete transformation results .
3.5 Prop erties of Integrated Flange Forgings [16]
3.5.1 Metall urgic al Homogeneity
Segregati on is of vital importance with respect to weldin g and
neutr on irradiation. This is particularly
true,
if the carbon
cont ent is bel ow 0.179s, the mechanical strength re guir ement s a re no t
satisfied, an d if the carbon content is higher than 0.2 4%, th e
weldabilit y is deteriorated due to hardening of the heat affected
zone ( H A Z ) .
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of mechanical properties is brought about by the homogeneity of
chemical composition and a n accurate and narrow range temperature
control at quenching and tempering.
(3) Directionality of mechanical properties
Fig.28 shows the directionality of tensile properties in three
directions (tangential, axial and radial) for KWU's shell flange.
There is no difference in both yield and tensile strength.
Slight differences in elonga tion and reduction of area are found,
but the difference is not remarkable.
Fig.29 shows the Charpy V-notch impact properties transition curves
for tangential, axial and r adial directions for KWU's shell flange
and COCKERILL's mono-blo ck vessel flange. Impact energy in all
three directions is high e noug h, and any directionality o f
properties is not significant.
The homogeneity in mecha nical properties described above is brought
about by decreasing the micro-segregation and non-metallic
inclusions. Thi s was accomplished by the deeper understanding of
steel making a nd using sui table forging techniques.
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4 WELD
4.1 Structu ral Weld [2]
4.1.1 Wel d Metal
(1) Requirement
Since the plat es and forgings must be welded together, it is obvio us
that the mechanical p roperty requirements of the welded region a nd
the asso ciated heat affected zone (HAZ) can be no less demanding
than those o f the base material itself, particularly as experience
shows that the most likely location for flaws is in the weld and
H A Z .
The ASME requirements for weld mechanical properties and
procedures are somewhat dispersed but appear mainly in Sections I , I
(NB-2300 an d
N B - 2 4 0 0 ) ,
and IX of the Code. They appear to be less
specific t han those for plates and forgings but there is a
requirement that all weldments should conform to all of the mini mum
mechanical proper ty specifications for the materials which are
joined by
welds.
This is clearly a desirable requirement.
Procedures and requirements for Charpy impact tests in particular
are given mainly in the ASME Code Section I . Minimum tensile and
notch toughne ss requirements which are specified in the ASME Co des
are summarized in Table 2. Their extension to cover the
deteri orati on o f properties under irradiation is implemented in 10
CFR 50 Append ix G and in the ASME
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Code Secti on I Appendix 6. Section I (Part C) of the ASME Code
includes general specifications for welding materials and methods
while Se ctio n IX deals with qualifying standards for welding
procedures as well as with the qualifications of the welders
themselves.
(2) Chemical composition
Deposit compositions of manual metal arc welds, associated with
nuclear ves sel s fabricated in Europe and the USA (Tables 3 and 4 ) ,
show that consumables have been employed which are capable of
alloying the deposit with Mh-Mo, Mn-Ni-Mo and Mn-Ni-Cr-Mo. Depo sit
strengths in the post-weld heat treated condition matching tha t o f
the base steel can be achieved by various combinations an d levels of
the elemen ts carbon, manganese, nickel, chromium and molybdenum, and.
this explai ns the variety of deposit analyses that can be foun d in
the litera ture. A similar situation is evident for submerged arc
welds where wires alloyed with Mn-Mo, Mn-Ni-Mo or Mn-Ni-Cr-Mo have
been used b y fabricators to achieve the required deposit strengt hs
(Tables 3 to 5 ) . However, the choice of flux is a very impo rtant
factor go ver ning the deposit composition of submerged ar c w e l d s ,
influencing in particular the carbon, silicon and manganese levels
and the im puri ty element levels of sulphur, phosphorus and o xyg en.
Fused f luxe s o f the calcium silicate type are favoured by m any
nuclear ves sel fabricators. These tend to give lower carbon levels
but higher silic on and oxygen levels in the deposit compared wit h
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agglomerated fluxes [22] , which are chemically mor e basic and
favoured by some European fabricator s, because the ass ociated
deposits generally have higher toughness . Most flux types will add
small amounts of phosphorus but the basi c fluxes a re capable of
lowering deposit sulphur levels, unlike the fused calicium si licate
fluxes.
For the beltline regions, where it is necessary to limit the copper
content to reduce the sensitivity to irradiatio n embrittlement, it
is necessary to depart from the practice of using copper-coated
electrodes, Hawthorne [21] has shown that copper contents belo w 0.1
wt9» are readily achieved i n sound wel ds providing the weldin g wi res
are protected from corrosion before use.
(3) Welding procedure
The main aims of evolving a satisfactory welding pro cedure ar e to
obtain the required mechanical properties in the weld, namely
strength and toughness, to produce a weld free from ultrasonic
'indications' which would require its repair and to avoid the
existence of cracks which would also require rep air but if remaining
undetected coul d act as the nucleus for fracture- proce sses .
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(4) Defects in the weld metal
The welding process can lead to a variety of defects in the weld
metal or adjacent HAZ of the parent materi al. Much is known about
the mechanisms of formation of these defects and how to avoid o r
minimise their occurrence by control of material composition and
fabrication procedure.
There are three forms of cracking that are potential problems in
weld deposits made in low alloy steels; solidification cracking,
reheat cracking and hydrogen induced cracking. There are no
published reports of incidences in nuclear vess el fabrication of the
first two forms , indicating that the consumables and pro cedures
normally selected in European and American nucle ar fabrication shops
have adequate resistance to these types of cracking. The
metallurgical factors controlling solidificat ion and reheat cracking
in weld deposits are reasonably well understood [23, 2 4 ] .
Hydrogen induced cracking can occasionally b e found
i n .
multi-pass
deposits in situations where the welding proce dure an d shop floor
storage and handling of consumables are not sufficiently controlled.
The cracks occur typically in arrays and are often transverse to the
line of the wel d and inclined at about 45* to the weld surface.
Sometimes referred to as chevr on cracks , the individual cracks hav e
a. zigzag appearance and c an b e transgranular or intergranular with
respect to the microstructure. They may be up to 40 mm in their
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longest dimens ion, but are often confined to a single weld pass and,
in these situations, are less than about 5 mm in length. The
factors controlling this form of cracking are similar to those for
hydro gen cracking in the HA Z, with cracking being more likely in
situatio ns w here the weld metal hydrogen content is high, the
restr aint is high and the weld deposit microstructure is
susce ptible, this normally means that the deposit is of high
hardness
[ 2 5 - 2 8 ] .
However, cracking can occur in weld metals at
significant ly lower hardnesses than would be associated with
crackin g in the HAZ . The relationship between weld hardness,
micro stru cture and susceptibility to cracking is the subject of
current resear ch but there is sufficient knowledge at present time
to s pecify adequate procedures for fabrication in order to avoid
cracking.
For nuclea r ves sels , weld metal hydrogen induced cracking must b e
regarde d as a potential pro blem in submerged arc and manual metal
arc weld depo sits . For welds of the former type, the type of
submer ged arc flux used is an important factor to be considered,
since there is evidence that agglomerated flux types have been
assoc iated w ith a greater tendency to weld metal hydrogen cracking
than fused fluxes [2 9]. Crack-free welds can be produced with
either fused or agglomerated fluxes but the latter type needs mor e
careful sho p floor control. As stated the choice of a fused flux,
to pro vide a goo d resistance to weld metal hydrogen cracking, bring s
with it the penalty of a generally lower weld metal notch toughness,
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due principally to the higher non-metallic inclusion content of such
deposits w he n compared with those made with agglomerated fluxes
which are mo re bas ic. It must also be noted that the level of
hydrogen in th e weld metal is partly determined by wire cleanliness
and thus wir e quality as well as flux quality is important.
4.1.2 Heat Affected Zone
(1) Defects in the weld metal
As for the wel d metal, there are three forms of cracking that are
potential pro blems in the HA Z, namely liquation cracking, r eheat
cracking and hydr ogen induced cracking.
(a) Liquatio n cracking
Liquation o r ho t cracking is a mode of intergranular cracking
occurring at elevated temperature in the initial welding thermal
cycle, i.e. befo re post-weld heat treatment. It is associated wi th
weak g rain bound ary zones o f reduced melting point material
containing enh anced concentrations of impurities particularly
sulphur, and o ccurs preferentially in localised regions of pos itive
segregation [30, 31 ]. Liquation cracking has been observed in the
HA Z of 22 NiMoC r 37 (similar to SA-508 Class 2) and in 20 MnMoNi 55
(similar to SA-5 08 Class 3 and SA-533B Class 1 ) but none has bee n
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reported for weld metal itself
[ 3 1 - 3 4 ] .
In examinatio n of 120
weldments from test plates, production weld prolongati ons and actual
components, about 3 0 * contained liquation crac ks w ith dimensions
typically of a few grain diameters (i.e.< 1 m m ) .
The control of liquation cracking is primar ily a question of control
of bulk and local purity, because other paramet ers such as welding
technique and heat input appear to be of s econ d or der significance
[ 3 0 ] .
Restrictions on copper, tin, phos phor us, sulphur and arsenic
bulk impurity content to the levels given in Ta ble 6 have been
proposed [33] together with a 'threshold crackin g criter ion
1
such
that cracking is likely to occur if two or mo re elements exceed
these value s. Whilst this criterion success fully characterises the
cracking susceptibility of the weldments ex amine d, it must b e
conceded that local regions of segregation may occasionally lead to
small isolated liquation cracks in material satis fying these bulk
compositional requirements. Liquation crac king susceptibility as a
phenomenon generic to weldable steels, is kno wn to decrease with
increasing manganese-sulphur ratios [ 35 ]. Alt hou gh all modern PWR
pressure vessel steels, whether of the SA-533B/S A-508 Class 3 or SA-
508 Class 2 type, have high manganese-sulphur rati os, the former
typically contains
1.6-1.8
times the manganese of the latter and
therefore for a fixed sulphur impurity conte nt it may be anticipated
to be less susceptible to this mode of cra ckin g.
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-
(b) Reheat cracking
Of the embrittlement and cracking phen omen a resu lting from the
fabrication of nuclear pressure vessel s, reheat cracking is
considered the most difficult to solve. Furthermor e, in spite of
extensive research of this phenomenon in PWR press ure vessel [9, 30 ,
33-49] and other steels, a complete understanding of all controlling
parameters has not yet been achieved, Howe ver, there is general
agreement in qualitative terms of the mechanis m involved and
susceptibility of individual classes of s teels ; for example SA-533B
Class 1 and SA-508 Class 3 although not immune are less susceptible
to this mode o f cracking than SA-508 Class 2.
Reheat cracks vary in size from a single grain (—20 ^ m ) to a
significant fraction of the weldment (10-im) [3 4, 3 5 ] . Cracking
occurs preferentially in regions of alloying and impurity element
segregation. Microcracks are restricted to the coarse-grained
unrefined regions of the HA Z whereas macrocra cks link coarse-grained
regions interconnected by partially refined regions of HA Z
microstr ucture. The positions and types of cracking found in the
extensive investigations by Kussmaul an d co-workers [30, 32-34] are
summarised in Fig. 30.
Reheat cracking is a high temperature gr ain boundary fracture
phenomenon occurring at temperatures bel ow about 700°c • The
occurrence is more likely in materials with coarse austenite gr ain
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size and rel atively high levels of the impurities together with a
high hardnes s before reheating or stress relieving and a resistance
to softening at elevated temperatures. Thus the extent of cracking
in heat affec ted zone is reduced by the presence of aluminium
nitride which induces a fine grain size and prevents grain growth by
minimising t he heat input into the weld to reduce the size of the
H A Z , and by restricting levels of those elements which form fine
dispersions of stable carbides and hence are responsible for
resistance to softening during the tempering heat treatment. For
the latter pur pos e it is important to achieve very low levels of
vanadium, zirconium and niobium.
Reheat crac king results wh en the relaxation strain exceeds the local
creep ducti lity of the material . It occurs during postweld heat
treatment (PWHT) when the welded structure is heated slowly from
room temper ature or the post-welding temperature (up to 300 °c) to a
temperature between 550 and 650 °c , held at this temperature for
several ho urs and slowly cooled to minimise further residual
stresses. Cracking can occur during heat-up or holding when the
instantaneous conditions o f residual stress, hardness, accumulated
strain, micro str uctur e and interfacial segregation of impurity
elements a re consistent wi th the requirements of a particular
failure mechanism.
The creep proc esse s occurr ing during relief of residual stresses are
sensitive t o small changes in alloy and impurity element composition
and to microstructure [31, 33, 34, 41, 44, 46,
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4 7 ,
5 0 ] . Whilst there is general agreement that SA-533B Class 1 and
SA-5 08 Cla ss 3 are less susceptible to reheat cracking than SA-508
Clas s 2 an d that impurity elements are deleterious, there is only a
broad quantitative consensus on the relative effects of different
impurities. From studies on simulated HAZs in experimental steels
contai ning single impurities, Brear and King [47] recommend that
individual elements should not exceed the limits given in Table 7
and the combined impurity element (nt.%) should be
P + 0.81 As + 1.18 S n + 1.49 Sb + 0.12 Cu + 0.195 S < 0.03
in order to avoid reheat cracking. Specimens such as those used in
this study, with simulated coarse-grained microstructure across the
whole cross-sectio n, in general , yield pessimistic results in
compariso n with those obtained from actual weldments. The results
also refer to specific test conditions and it is not clear how the y
relate quantitatively to the occurrence of reheat cracking in
service . Kussmaul et al. report an increasing tendency for cracking
with increasing contents of phosphorus, sulphur, copper, arsenic,
aluminium, nitro gen, molybdenum and cobalt [33 ]. They also propose
a 'threshold value
1
criterio n such that reheat cracking occurs if
two or mo re elements exceed the limits given in Table 6. More
recently [3 8] , examination of reheat cracking in commercial casts of
SA-533B C las s 1 and SA-508 Class 2 steels has indicated that
increasi ng the chromium content is deleter ious. Similarly the
importance of chromium and other strong carbide-forming elements
such as m olybdenum, vanadium, titanium and niobium in prompting
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rehea t cracking in other low alloy steels has bee n recog nised [50,
5 1 ] .
(c) Hydro gen cracking
Hydro gen cracking is a potential source for d efects associated wi th
the HAZ of structural weld. This is a brittle crac king mechanism
occurring below about 200°c. The phenomenon is associat ed
particularly with high strength steels in the as-welded condition
wher e hydrog en has been introduced during welding and high levels of
stress rem ain [53]. Although no failures or large defects in PWR
plant have be en reported to be caused by hydrogen cracking, recent
wor k h as demonstrated that regions of alloy segregati on in PWR
pressu re vessels and other steels are more susceptib le to hydrog en
cracking than the matrix [ 4 9 , 5 4 - 5 6 ] . Liquation cracks , also
occurring in segregated material, are ideally suited to act as
stress concentrators for subsequent hydrogen c rackin g.
The critical concentration of hydrogen below w hich crack initiation
will not o ccur is not known, although values as low as 1.3 ppm hav e
bee n suggested [56]. A low hydrogen content in plate s and forgings
is achieved by vacuum degassing prior to casting (2-3 ppm) together
wit h one or sore heat treatments during fabrication (1 p pm ). O n
welding, the local hydrogen concentration of the weldmen t will
increase.
Wit h good welding practice, as specified in procedures
for fabrication of nuclear vessels, a concentration of about 5 pp m
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ca n be expected. Immediately after welding the HAZ is susceptible
to hydro gen cracking and therefore should be maintained at a n
elevated temperature until sufficient hydrogen has diffused
away.Tests have shown that no cracking occurs in SA-508 Class 3
weldments provided that either a 200 °c preheat and post-heat
temperature is employed or a lower preheat temperature combined w ith
a post-heating cycle after welding is used [3 7]. Further reductions
in hydrogen concentration will occur during subsequent stress relie f
annealing.
4.2 Cladding
(1) Material and welding procedure [2]
Th e inner surfac e of the vessei is clad with a corrosion resistant
layer by continuously melt ing cladding material onto the vessel
surface to produce a fusion weld. The method is mechanised for the
singl e curvatur e surfaces in order to provide a layer of constant
thickness. Certain regions of double curvature, however, must be
clad manually. Two types of feed material are used; a type 309/308
austenitic stainless steel which is used for cladding the ferritic
stee l of the ma in pressure vessel and Inconel overlay which is use d
o n penetrations , core support pads and on the faces of the nozzl es.
Detai ls of established m etho ds of clad welding, electrode and flux
compositio ns and operating conditions are given in the review of
overlay welding by Gooch [5 7] . A recent process has been developed
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reportedly capable of depositing high quality austenitic cladding up
to 300 nun with [5 8] .
(2) Defe cts associated wit h cladding [2]
There are several differences between structura l weld ing and
cladding which are relevant to the mechani sms of formatio n of
welding defects. In the case of cladding the depos it is mainly
austenitic rather than ferritic, and therefo re poss esses a higher
coefficient of thermal expansion and greater hig h temperature
strength, both factors influencing residual stre ss fo rmation and
stress relaxation behaviour. Also heat input is high during
cladding thus promoting residual stresses and large heat affected
zones.
Howeve r, with the exception of the ins ide of nozzles the
cladding process is a low-restraint weld co nfigur ation and th erefore
less severe with respect to long range resid ual stress fo rmation.
Hydrogen ha s a higher solubility but lower diffusivity in austenitic
than in ferritic steels. Consequently, the austenitic cladding can
retain hydrogen which may subsequently pass into the base material.
Most repor ted defects associated with the cladd ing of PWR p ressu re
vessels are related to the HAZ in the ferritic steel below the
cladding. However, defects can occur both in the cladding itself
and along the cladding/ferrite interface. Goo ch reports three
instances of cracking in the cladding duri ng fabricatio n due to
failure to achieve the desirable compos ition an d microstructur e
[ 5 7 ] .
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(a) Underclad reheat cracking [2]
A potential problem associated with cladding of r eactor pressure
vessels is the formation of underclad cracks . The first report of
defects in the HAZ beneath austenitic cladding in nuclear plan t w as
in 1970 [59 ]. The Welding Research Council undertook a
comprehensive review of the phenomenon and Vinckier and Pense
reported this work in 1974 [ 9] . Cracks were found exclusively alon g
prior austenite grain boundaries with sizes varying from a minimum
of 0.2 mm in depth and length to a maximum of 10 mm length and 3 mm
depth. In mor e recent reviews [31, 44] the maximum depth is
reported as 4 mm. The cracks exist in a region which is somewhat
difficult to examine by conventional ultrasonic testing techniques
because of the proximity of the cladding and the surface. Cracks
have been revealed by stripping the cladding and using surface cr ack
detection method s. Metal lographic examination showed that cracks
wer e in the coarse-grained region of the HAZ which ha d been fully
austenitised by the first cladding deposit and then heated to ju st
below the austenitisation temperature, i.e. 600-700°c, by the
subsequent adjacent cladding deposit. The susceptible region is
under the highes t residual tensile stress immediately after welding .
The direction of cracking wa s usually between 45 and 90* to th e
direction of welding. Fig . 31 illustrates the positi on of the
cracks.
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Vinckier an d Pe nse concluded that pressur e vessel steels
manufactur ed t o the different specifications have different
susceptibil ities to underclad cracking. Out of 96 reports showing
26 cases o f u nderclad cracking, 25 wer e in SA-508 Class 2, one in SA-
508 Class 3 and no cases were reported for SA-533B Class 1. High
heat input duri ng cladding resulted in underclad cracking in SA-508
Class 2 but not in the other steels. This pattern of behaviour was
confirmed b y other reviews in the period 1974-1978 [31, 44 , 60 , 6 1 ] .
The cr ackin g reported was all attributed to reheat cracking
occasi onally augmented by liquation cracking.
Figs.
32 an d 33 show reheat cracking susceptible areas and methods
to avoid reheat cracking by refining heat affected coarse grain
zones [6 2] .
(b) Under clad hydrogen cracking [54]
Hydrogen cr ackin g is one of the most important problems for the
integrity of steel structures and man y studies on this subject have
been extens ively carried out until now . The dominant factors for
the hydr ogen cracking are summarized as follows:
- existe nce of diffusible hydrogen
- existence of tensile stress or strain
- exist ence of hydrogen embrittlement susceptible
microstrueture
- low temperature condition s (below 150°c)
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The hyd rog en crack initiates when all four factors are
simult aneou sly satisfied. In the case of welding between low alloy
ferriti c m et als , the weld joint becomes quite susceptible to the
hydr ogen crack ing and almost all the studies have been conducted o n
this su bje ct. On the other hand, weld joints of ferritic metal and
austen itic wel d metal are considered to be resistant to hydrogen
crac king b ecau se of the high hydrogen solubility, low hydrogen
diff usio n r at e in the weld metal and sufficient capacity of
relax ation o f welding induced strain. Nevertheless, the hydrogen
cracks in t he HA Z under austenitic stainless steel overlay wer e
recent ly repo rted o n the tube sheet forging of a steam generator in
a light wa ter reactor [55, 6 3 ] . The references pointed out that the
cracks ma in ly initiate in the zones of segregation.
Th e ex ist enc e of susceptible microstructure to hydrogen
embr ittl emen t is also an important factor influencing the initiation
of hydrogen cracks.
Fou r vit al conditions to initiate the hydrogen cracking have been
examined independently for the heat affected zone under the
auste nitic stainless steel overlay. From the results, it is
concluded that :
- 2 t o 4 ppm hydrogen diffuses from the austenitic weld
me ta l into the bas e metal by the «p-transformation of th e
ba se metal due to the welding
- he at affected zo ne is quite hydrogen embrittlement
susceptibl e and and the embrittlement is
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especially noticeable at room temperature. The r emarka ble
hydrogen embrittlement occurs at hydrogen content of 1.5
ppm and higher
- segregation, which is difficult to avoid in large
forgings at the present time, is the most susce ptible to
hydro gen embrittlement
- maximum magnitude of residual stress amo unt to 500 MP a an d
the magnitude is sufficient to initiate the h ydro gen
cracking
From t he above facts, it is quantitatively proved that hy dro gen
cracking occu rs even in the weld joint between the austen itic we ld
metal an d ferritic metal.
The preventive measures against hydrogen cracking a re s ummarized as
follows. On the assumption that the materials and desi gn of
component ar e not changed for this purpose, the exist ence of
hydrogen embrittlement susceptible microstructure and restr aint
condition of weld joint cannot be avoided. On the other hand, t he
large decrease s in residual stresses under the weld over lay cannot
be expected by the conventional soaking treatm ents. Hen ce, the
countermeasur es against the hydrogen cracking become s as fo llo ws.
- elimination of hydrogen in the heat affected zone
- to avoid the low temperature conditions un der the hydr oge n
abs orb ed conditions of heat affected zone
Therefo re, it is recommended that the preheating should be
maintained a t least during 1st and 2nd layer welding and until pos t
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wel d heat treatment or soaking treatment. The soaking treatment at
about 250°c is undoubtedly effective to avoid the hydro gen
cracking. By these countermeasures, hydr ogen cracking in the HAZ of
heavy forging with stainless steel overlay cla d can be complet ely
avoided.
4.3 Mechanical Properties of Weld Metal and Heat Affected Zo ne
(HAZ) [2]
Tensile and Charpy V-notch impact properties for the European weld
metals are comparable with the base metal data . It would appear
that adequate low temperature notch toughness and comparable tensile
ductility can be achieved in weld metals even when the weld metal
yield strength is approximately 10-20& higher than the nominal mean
value o f 470 MPa for SA-533 Grade B Class l/SA-508 Class 3 ba se
materials. Data for base metals, weld metals and HAZs which were
tested in the EPRI programmes [20] are collated in Tables 8, 4 and 9
to 1 2. Dat a from Japan which provide some indication of the
improvement in upper shelf notch toughness a nd also in the ductile -
brittle transition temperature that can be o btained using nar row gap
welding processes [64, 6 5 ].
From the results examined for both weld metal s and HAZs, which are
rather limited in some cases, it can be concluded that it is
possible to achieve mechanical properties in weld metals and he at
affected zones
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that are at least as good as those of plates and forgings. It is
believed that the weld metal rather than the HAZ will govern
acceptance. By correct selection of welding consumables/
parameters, weld metal properties can be achieved which are well in
excess of the minimum property values specified by the ASME
C o d e s .
The upper shel f notch toughness of weld metals should also be
reasonably high , but cannot be guaranteed to match the high notch
toughness values of the very high quality steel now available from a
number of sour ces. Recent developments in the production of plates
and forgings , and to a lesser extent w e l d s , have resulted in an
improvement in properties, beyond the minimum values required by the
A S M E C o d e s , and therefore beyond the material properties upon which
p r e v i o us P W R v e s s e l p r o d u c t i o n h a s b e e n b a s e d .
5 A d v a nc e d D e s i g n d u e t o O p t im i z ed M a t e r i a l
The design of the RPV for the light water reacto r (LWR) tends to
m i n i m i ze t h e w e l d s e a m s , which reduces the period of in-service
inspection (ISI) together with easier perfo rmance of ISI. This
tendency reg uired the more integrated and larger pa rts for nuclear
steam supply system (NSSS) components.
It was said t hat weld seams can be reduced to 70 percent for boiling
water reactor pres sure vessels (BWRPVs) and 25 percent for
pressurized wate r reactor pressure vessels (PWRPVs) when compared to
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conventional desig ns, by the use of the large forgings and plates
available at present in the world [6 6] . Typical layouts for these
designs of nu clear pressure vessels are as shown in Fig.34 for the
P W R a n d F i g . 3 5 f o r B W R.
The seamles s forged shells for the BWRPV and PWRPV, as well as the
vessel flang e integral with nozzle belt shell in PWRPV, are
significantly advantageous from the standpoint of design,
fabrication and inspection.
The nume rou s seamless for ged shells in BWRPV and PWRPV have already
b e e n r e a l i z e d a s s h o w n i n F i g s . 3 6 a n d 3 7 [ 6 7 ] , r e sp ec ti ve ly , u s i n g
the advan ced technology for the manufacture of heavy steel forgings.
One-piece fo rg ed shell flanges weighing 165 tons for KWU type
4-loop
P W R P V m a d e f r o m 4 0 0 t o n in g o t s s h o w n i n
Fig.
1 7 h a v e b e e n ~
s u c c e s s f u l l y d e v e l o p ed [ 1 3 ] . F u r t h er m o r e , o n t h e b a s is of m u c h
manufa cturi ng experience of one-piece shell flange for KWU, mo no -
block ves sel flange of WEC t ype 157 " (3988 mm ) PWRPV, combined
v e s s e l f l a n g e w i t h n o z z l e b e l t s h el l a n d o n w h i c h t he s e t -o n t y p e
m a i n c o o l a n t n o z z l e s a r e w e l d e d a s s h o wn i n
F i g .
1 8, was manufactured
s u c c e s s f u l l y u s i n g a 5 0 0 t o n i ng ot [6 8 - 7 1 ] .
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REFERENCES
[1] ON OD ER A, S., Tetsu To Hagane, Current Steels for Nuclear Pressure Ves sels, J.
ISIJ (Iron and Steel Institute of Japan), 67 (1981), P880
[2] Second Rep ort on Assessment of the Integrity of PW R Pressure Ves sels, UK.AEA
(Mar. 1982)
[3] STEE LE, L.E., Neutron Irradiation Emb rittlement of Reactor Pressure Vessel
Steels, IAEA (1975)
[4] AN DO , Y., Private Com munication to Mr. Steele, L.E. (21 No v. and 10 Dec .
1973)
[5] HAS HIMO TO, U., KIHARA, H., AND O, Y., Welding Problems Associated with
Construction of Nuclear Power Stations in Japan, 1964 Annual Assembly of IIW
(International Institute of Welding), Prague, Czechoslovakia (30 Jun. - 2 Jul. 1964).
Proceedings
[6] RA O, S., Private Com munication to Mr. Steele, L.E. (14 No v. 1973)
[7] PA CH UR , D., Private Com munication to Mr. Steele, L.E. (11 No v. 1973)
-
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- 113 -
[8] DE BR AY , W., CERJAK , H., Werkstoffeigenschaften des Stahles 22 NiM oC r 37
fur Reaktorkomponenten, VGB-Werkstofftagung (1971)
[9] VIN CK IER, A.G., PENSE, A.W ., A Review of Underclad Crack ing in Pressure
Vessel Components, WRC Bulletin 197 (Aug. 1974)
[10] CER JAK , H., DEB RAY , W., PAPOU SCHE K, F., Eigenschaften des Stahles 20
MnMoNi 55 fur Kernreaktor-Komponenten, VGB-Konferenz, Werkstoffe und
Schw eisstechnik in Kraftwerk 1976 (197 6), Diisseldorf
[11] KUS SMA UL, L., EWA LD, J., MAIER, G., SCHE LLHA MM ER, W., Enhancement
of the Quality of the Reactor Pressure Vessel Used in Light Water Power Plants by
Advanced Material, Fabrication and Testing Technologies (Aug. 1977), San
Francisco
[12] ON OD ERA , S., FUJIOKA , K., TSUK ADA , H., SUZ UK I, K., Ma terial
Specifications towards more Reliable Fabrication of Nuclear Pressure Vessels, The
2nd Joint Symposium of TUV Rheinland and B & W/USA (28 Sept. 1978), Koln.
Preprint
[13] DEB RAY , W., CERJAK, H., ONODERA , S., TSUK ADA , H., SUZ UKI, K., Large
Nozzle Belt Forgings for PWR 1200 MWe, 1st European Nuclear Conference
(April 1975), Paris. Preprint
-
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45/95
- 114 -
[14] U.S. NRC Regu latory Guide 1.99, Rev. 1, Effect of Residual Element on Predicted
Radiation Damage to Reactor Vessel Materials (Apr. 1977)
[15] JSW (Japan Steel W orks), In-C omp any Document (1 June 1982)
[16] ONODERA , S., FUJIOK A, K., TSUK ADA , H., SUZUK I, K., Advantages in
Application of Integrated Flange Forgings for Reactor Vessels; The 3rd MPA
Seminar (15 Sept. 1977), Stuttgart, MPA, University of Stuttgart, Germany, 1978
[17] COU PETTE, W., Der Einfluss der Seigerung und Verschm iedung auf die
Festigkeitseigenschaften grosser Schmiedestucke aus Stahl, Stahl und Eisen, 61
(1941),
P1013
[18] WEL FLE, K., BITT ER SM AN N, H., Flockenfreigliihen mit
Zwischenstufenumwandlung, Neue Hiitte, 11 (1966), P730
[19] BRU CKN ER, E., et al., The Properties of Various Weld Metals for Reactor
Components, DVS Berichte No. 52 Welding in Nuclear Engineering 1978 67-73.
V.E. Riecansky, Technical Translation No. VR/1318/78, Cambridge
[20] Nuclear Pressure Vessel Steel Data Base . EPR I NP9 33 Proj. 886-1 (December
1978)
[21] HAW THO RNE , J.R., W elding J., 51 369S (1972)
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46/95
-
115
-
[22] DAVIS, M.L.E., BAILEY, N., How Submerged Arc Flux Composition
Influences Element Transfer, Proceedings of Conference on Weld
Pool Ch emis try an d Metallurgy, The Weldi ng Institute (April
1980)
[23] BAILEY, N., JONES, S.B., Solidification Cracking of Ferritic
Steels dur ing Submerged Arc Welding, The Welding Institute
(1977)
[24] BATTE, A. D, MURPHY, M.C., Reheat Cracking in 2 Cr/Mo Weld
Metal : Influence of Residual Elements and Microstructure,
Metals Technology, Vol.6, No.2 (February
1 9 7 9 ) ,
P62
[25] HART, P.H.M., Weld Meta l Hydrogen Cracking, Welding Institute
Research Bulletin (November
1 9 7 8 ) .
[26] KEV ILLE , B.R., An Investigation to Determine the Mechanism
Involved in the Formatio n and Propagation of Chevron Cracks in
Submer ged A rc Weld ment s, Welding Research International 6 (6)
(1976)
[27] MOTA, J.M.F.,
APPS,
R.L., JUBB, J.E.M., Chevron Cracking in
Manual Met al Arc Welding, Proceedings, Trends in Consumables
and Ste el s for Weldin g, The Welding Institute, (November
1 9 7 8 ) , London
-
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-
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- 117 -
[33] KUSSMAUL, K., EWALD, J., et al., Enhancement of the Quality of the Reactor
Pressure Vessel Used
in
Light Water Power Plants
by
Advanced Material
Fabrication and Testing Technologies; 4th Int. Conf. on Structural Mechanics in
Reactor Technology; Paper Gl/3 (August 1977), San Francisco
[34] KUSSMAUL, K., EWALD, J., Assessment of Toughness and Cracking in the Heat
Affected Zone
of
Light Water Reactor Components;
3rd Int. Conf. on
Pressure
Vessel Tech., Part II, P.627-646, ASME (1977), Tokyo
[35] Solidification Cracking of Ferritic Steels during Submerged Arc Welding; The
Welding Institute (1977)
[36] BREAR, J.M. and KING, B.L., Phil. T rans. Roy. Soc, London A295 (1980)
[37] COMON, J., A508 Class 3 Forgings for Pressure Vessels; 3rd Int. Conf. on
Pressure Vessel Technology, Part
II,
P.957-970, (April 1977), Tokyo, Japan.
Atomic Energy Society of Japan, 1977
[38] McMAHON, C.J., DOBBS, R.J., GENTNER, D.H., Stress Relief Cracking in
MnMoNi and MnMoNiCr Pressure Vessel Steels, Mat. Sci. and Eng. 37 (1979),
P. 176-186
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- 120 -
[49] TSUKAD A, H., TAN AKA , Y., OHN ISHI, K., Temperature Dependence of
Hydrogen Embrittlement of Weld Thermal Cycled Material; Japanese Welding Jnl.
[50] VIEILL AR D- BA RO N, B ., Deve lopm ent of the Production of Special Steels for
Nuclear Industries, Materiaux et Techniques (January 1977), P321-337
[51] ITO, Y., NA KA NIS HI, M ., IIW (International Institute of W elding), Doc. X- 668-
72
[52] NAK AMU RA, N., NA IKI, T., OKA BAY ASHI, H., Fracture in the Process of
Stress Relexation under Constant Strain; Proc. First Int. Conf. on Fracture (1965),
Sendai, Japan, P863-878
[53] JPVRG Report N o. 2, Tem per Em brittlement and Hydrogen Embrittlement in
Pressure Vessel Steels; Iron and Steel Inst. of japan (May 1979)
[54] OHN ISHI, K., TSUK ADA , H., SUZUK I, K., MU RAI, H., KAG A, H.,
KUSUHASHI, M., OGAWA, T., TANAKA, Y., Study on Hydrogen Induced
Cracking of Heavy Forgings Overlaid by Stainless Steel; The Special Topical
Meeting, Metal Performance in Nuclear Steam Generators; ANS (6-9 Oct. 1980),
St. Petersburg Beach, Florida
-
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52/95
- 121 -
[55] CER JAK, H., SC HM IDT , J., and LOH NB ERG , R., Correlation between
Segregation and Cold-Cracking; 4th MPA Seminar on Safety of Pressurized LWR
Containments (October 1978), Risley, Tran.4253
[56] WID AR T, J., Private Com mun ication
[57] GO OC H, T.G., Re view of Overlay Welding Procedures for Light Water Reactor
Pressure Vessels; Welding Institute Document 3455/1/75 (1976)
[58] Kawasaki Steel Corp oration, Mag lay; The Ne w Surfacing Process with Electroslag
Welding using Wide Strip Electrode
[59] W YLIE , R.D., Rep ort of PVR C Task Group on Underclad Cracking, HSST (Heavy
Section Steel Technology); 6th Annual Information Meeting (April 1972), ORNL
Conf. 720468
[60] DO LB Y, R.E., and SAU ND ER S, G.G., Underclad Cracking in Nuclear Vessel
Steels, Part I, Occu rrence and Mech anism of Cracking. Met. Constr. (December
1977),
P562-566
[61] DO LB Y, R.E. and SAU ND ER S, G.G., Underclad Cracking in Nuclear Vessel
Steels, Part II, Detection and Control of Underclad Cracking. Metal Constr.,
(January 1978), P20-24
[62] VIG NE S, A., Private Com mun ication to Dr. Onodera, S. (Jan. 1983)
[63] Cracks in French Pressu re Vessels Pose N o Danger; Nuclear Engineering
International (Jan. 1980), P27
-
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53/95
- 122 -
[64] SAW ADA , S., OHT A, M., NISHIOKA, K., HORI, M., KAW AHA RA, M.,
YAMASAKI, H., Application of Narrow-Gap GMA Welding Process to Nuclear
Reactor Pressure Vessels; presented 4th Int. Conf. on Pressure Vessel Technology,
PI 13-1 19, Vol. II, I. Mech. E. (May 1980), London
[65] MO RIG AK I, O., et al., Submerged Arc Narrow-G ap Welding Process with One
Run per Layer Technique for Heavy Sections; Nippon Steel Welding Products and
Engineering Co. Ltd. (February 1979(, IIW (International Institute of Welding),
Doc. XII-A-168-79
[66]
ISHIKA W A, K., AN DO , E., Experience Leads to Major Advances in Reactor
Pressure Vessel Design; Nuclear Engineering International (July 1977)
[67] JSW (Japan Steel Wo rks, Ltd.); In-Company Document
[68] ONO DERA , S., MO RITANI, H., TSUCHIYA, K., TSUKADA , H.
5
WIDART, J.,
SCAILTEUR, A., Mono-Block Vessel Flange Forgings for PWRPV 1000 MWe;
the 8th International Forgemasters Meeting (October 1977), Kyoto
[69] CA M BIEN , R.B ., New Design of PWR Reactor Vessel Using Large Forging ;
ASME Conference (September 1976), Mexico City
[70] RE YN EN , J., W IND T, P.DE., WIDA RT, J., et al., A Novel Design for LW R
Pressure Vessel Nozzles and Corresponding Stress Analysis; 3rd International
Conference on Pressure Vessel Technology (April 1977), Tokyo, Japan. Atom ic
Energy Society of Japan, 1977
[71] W IDA RT, J., Design of Reactor Pressure Vessel Considering Easier I.S.I.; IAEA
Technical Committee (April 1977), Kobe
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- 124 -
TABLE 2
M in im um M e c h a n i c a l P r o p e r t i e s S p e c i f i e d i n t h e ASME C o de s ( 2 )
A 53 3B 1 P l a t e s A 50 8 3 F o r g i n g s
T e n s i l e 2 0 *C 3 5 0 ' C 2 0
8
C 350°C
Yield stress (MPa)
U l t i m a t e t e n s i l e
stress (MPa)
El ( in 50 mm) %
R of A %
Charpy Impact
Energy (J)
Lateral expansion (mm)
Minimum ave. value (+)
of three specimens
Minimum value of one
specimen
345 285(*)
552 527
18
-
68 J a t
0 .8 9 mm
X
X
34 5
550
18
38
RT
ND T
+ 33"
at RT
ND T
+
41 J a t
34 J a t
285
-
-
C
33°C
4.4°C
4.4*C
("*) non-mandatory
not more than
this value
to be specified by purchaser
(*) no t more than one specimen from a se t may fa ll below
this value
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- 126 -
As-Deposited
Heat C
Chemical
Mn
Analyses
P
TABLE
of EPR1
iwt.
S
4
. ' .SO Heat
1
" ( 2 0 )
S i Ni
Weld Metal
•
Cr
s in
Mo
A533B
Y
Cl 1
Cu
MMA
Welds
P
Q
R
S
T
U
S/A Welds
V
W
X
Y
Base
~B~~
L
N
0.100
0.100
0.100
0.09
0.04
0.050
0.150
0.14
0.150
0.13
0.25
0.21
0.24
1.000
1.110
1.100
1.03
1.02
0.150
1.38
1.19
1.280
1.25
1.41
1.34
1.30
0.005
0.007
0.006
0.005
0.017
0.016
0.008
0.01
0.011
0.011
0.008
.0.012
0.009
0.010
0.010
0.010
0.01
0.022
0.024
0.009
0.009
0.010
0.010
0.014
0.019
0.013
0.390
0.400
0.400
0.39
0.49
0.520
0.16
0.19
0.200
0.18
0.260
0.230
0.240
0.880
1.060
1.000
0.95
0.95
0.940
0.13
0.10
0.190
0.10
0.46
0.44
0.46
0.010
0.010
0.010
0.01
0.01
0.010
0.04
0.09
0.080
0.09
0.11
0.07
0.11
0.290
0.340
0.330
0.32
0.53
0.540
0.60
0.54
0.540
0.53
0.49
0.53
0.53
0
0
0
0
0
0
0
0
0
0
0
0
0
.004
.006
.005
.006
.014
.014
.007
.005
.005
.005
.003
.004
.002
0.020
0.020
0.020
0.020
0.02
0.030
0.04
0.12
0.110
0.20
0.12
0.10
0.08
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TABLE 5
(
C h e m ic a l c o m p o s i t io n , o f f i l l e r me t a l
as
s p e c i f i e d ,
as
r e c e i v e d
and as
d e p o s i t e d ,
t o g e t h e r w i t h t h a t
o f t h e
base meta l (SA533B)
( 21 )
Ch e m i c a l c o m p o s i t i o n , wt .%
C
Mn P S Si Ni Cr Mo Cu V A l As Sn Sb
F i l l e r m e t a l :
s p e c 1 f 1 c a t i o n
a
. 15 1 .80 LAP LAP 0.1 0 .55 .10 .45 LAP .02 .05 LAP LAP LAP
7?0~ 2 TT0" OTOTO OTOTO Tma x)
T7 F (max) j5o~ 0T W
(inaTx)
( i a x )
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TABLE. .9
Tensile P roperties
o f
EPRI Base Metals and Melds
( 2 0 )
M a t e r ia l
Tensile M-ft weld i n M f l weld in S /A weld i n M A IIAZ i n S/AI1AZ in
Prope rly A533B1 base ( T ) AS3301 ( L ) A53301 ( T ) A533B1 ( L ) A53301 ( L ) A533U1 (L )
Y i e l d s t r e s s
(f-Pa)
UTS
E l o n g a t i o n
R o f A
T'C
-24
200
-2 0
288
- 2 0
288
- 2 0
288
n
13
tl
II
•1
It
tl
' II
II
X
435
386
592
578
27.7
23.3
64.1
56 .4
lsd
2 9 . 9
25 .0
27.3
2 4 . 4
2. 0
2. 1
3.2
5.5
T'C
43
208
43
288
43
288
43
288
n 3f
4 478
" 437
" 577
" 570
11
3 0 . 3
" 27.4
" 74.6
" 6 9 . 0
ls d
4. 5
27.1
4.2
9.7
1.7
111
1.1
3.1
T'C
23
200
23
288
23
288
23
288
n
x
2 400,
556
11
4 3 4 ,
474
11
607,
632
" 500,
599
11
15 .5 ,
26 .2
" 19.0 ,
22.5
" 3 6 . 7 ,
65 3
11
6 6 . 7 ,
67.3
T'C
40
200
40
288
40
288
40
288
n
4
II
II
"
II
II
II
H
J
537
470
620
596
25.6
2 3 . 9
72 .1
65 .1
lsd
14.1
19.3
8.8
23.3
0. 5
1. 5
0.7
1.8
T'C
23/43
288
23/43
288
23/43
288
23/43
288
n
3
H
"
H
II
It
II
H
X
444
412
598
591
25.3
21.6
69.7
59.1
1st
4 7 . 9
3 9 . 8
4 4 . 8
36 .0
1.2
1.6
3.4
10.2
T'C
60
208
60
280
GO
288
60
200
n
1
it
"
M
II
It
II
II
x"
403
395
555
590
22
22
60.5
61.8
A508-2 base
( L )
m veld
in S /A
waid
i n
A508-2
( L )
A508-2
( L )
Cfy (M>a)
o \ (M>a)
El {%)
R o f A
(%)
•
. T'C
24
288
24
288
24
208
24
2 2 0
n
x
5 440
" 402
11
595
". 585
" 27.4
11
24.3
" 70.0
" 66.6
ls d
3 3 . 6
26 .7
2 3 . 5
21.2
1.0
0. 5
1.1
1.65
n S
5 526
" 443
" 611
" 567
" 27.0
11
22.9
" 71.6
11
64.4
1st
50.5
20.6
52.9
31 .1
1.0
1.6
1.4
4.5
n S
5 507
" 434
" 619
" 5(30
" 25 .3
" 21 .0
11
67.4
" 62.4
ls d
47
30.1
36 .2
32.5
1.7
1.7
4 .3
5. 5
n
» n o . o f
heats
x - mean property value
(L )
"
longitudinal orientation
(T ) « transverse orien tation
-
8/17/2019 Nuclear Pressurized Water Reactor (PWR) - Mechannical design materials
62/95
-
8/17/2019 Nuclear Pressurized Water Reactor (PWR) - Mechannical design materials
63/95
-
8/17/2019 Nuclear Pressurized Water Reactor (PWR) - Mechannical design materials
64/95
- 1 3 3 -
Charpy Impact and
Electrode
/Flux
E8O15.C3
u
II
E8O18.C3
n
MnMoNi/
Linde 8 0
HnMoNi/
Linde 0 0 9 1
Heat
Input
W/m)
Manual
f
TABLE
1 1
RT
MD T
D a t a
f o r
W e l d s
i n
A 5 0 8 - 2
(Effect
of
Heat Input)
Heat,
Weld
RT
NDT
CO
total Arc
0 . 8 G
1.2 I
2 . 2 K
0 . 7 H
2 . 3 J
Submerged Ar c
3 . 2
3 . 6
4 . 0
2 . 8
4 . 0
P
0
L
N
M
- 7 3
- 6 2
- 6 2
- 4 0
- 5 1
- 1 8
- 1 8
- 9
- 5 1
- 6 2
Metal
To C
CO
- 3 8
- 3 7
-