SCK-CEN Contribution to the IAEA Round Robin Exercise on WWER-440 RPV Weld Material: Irradiation, Annealing, and Re-Embrittlement
Mechanical properties of WWER-440 RPV weld joints have been studied with account of different states of the material: baseline (unirradiated), irradiated up to the average fast neutron fluence of 6.5*10^19 n/cm^2, irradiated and eventually annealed, re-irradiated with the accumulated fast neutron fl...
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| Опубліковано в: : | Проблемы прочности |
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| Дата: | 2004 |
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Інститут проблем міцності ім. Г.С. Писаренко НАН України
2004
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| Назва журналу: | Digital Library of Periodicals of National Academy of Sciences of Ukraine |
| Цитувати: | SCK-CEN Contribution to the IAEA Round Robin Exercise on
 WWER-440 RPV Weld Material: Irradiation, Annealing, and
 Re-Embrittlement / E. Lucon, E. van Walle, M. Scibetta, R. Chaouadi, M. Weber // Проблемы прочности. — 2004. — № 1. — С. 30-47. — Бібліогр.: 22 назв. — англ. |
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Digital Library of Periodicals of National Academy of Sciences of Ukraine| _version_ | 1860268224789086208 |
|---|---|
| author | Lucon, E. Walle, E. van Scibetta, M. Chaouadi, R. Weber, M. |
| author_facet | Lucon, E. Walle, E. van Scibetta, M. Chaouadi, R. Weber, M. |
| citation_txt | SCK-CEN Contribution to the IAEA Round Robin Exercise on
 WWER-440 RPV Weld Material: Irradiation, Annealing, and
 Re-Embrittlement / E. Lucon, E. van Walle, M. Scibetta, R. Chaouadi, M. Weber // Проблемы прочности. — 2004. — № 1. — С. 30-47. — Бібліогр.: 22 назв. — англ. |
| collection | DSpace DC |
| container_title | Проблемы прочности |
| description | Mechanical properties of WWER-440 RPV weld joints have been studied with account of different states of the material: baseline (unirradiated), irradiated up to the average fast neutron fluence of 6.5*10^19 n/cm^2, irradiated and eventually annealed, re-irradiated with the accumulated fast neutron fleunce 1.3*10^20 n/cm^2. For each state of the material, tensile, impact fracture, and fracture toughness tests were performed with the use of Charpy (standard, reconstituted, and pre-cracked) specimens.
Исследованы механические свойства материалов сварных соединений корпусов реакторов ВВЭР-440 с учетом различных состояний: исходного (неохрупченного), после радиационного охрупчивания со средней плотностью потока нейтронов до 6,5*10^19 нейтр/см^2, охрупченного с последующим отжигом, повторно охрупченного с флюэнсом частиц 1,3*10^20 нейтр/см^2. Для каждого состояния материала проведены испытания на растяжение, ударное разрушение и трещиностойкость с использованием образцов Шарпи (стандартные, восстановленные, с предварительно нанесенной трещиной).
Досліджуються механічні характеристики зварних з ’єднань корпусів реакторів BBEP-440 з урахуванням різного стану: початкового (неокрихчувано- го), після радіаційного окрихчування з середньою густиною потоку нейтронів до 6,5*10^19 нейтр/см^2, окрихчуваного з наступним відпалом, повторно окрихчуваного з флюенсом частинок 1,3*10^20 нейтр/см^2. Для кожного стану матеріалу проведено дослідження на розтяг, ударне руйнування та тріщиностійкість із використанням зразків Шарпі (стандартні, відновлені та з попередньо нанесеною тріщиною).
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| first_indexed | 2025-12-07T19:03:15Z |
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| fulltext |
UDC 539.4
SCK-CEN Contribution to the IAEA Round Robin Exercise on
WWER-440 RPV Weld Material: Irradiation, Annealing, and
Re-Embrittlement
E. Lucon, E. van W alle, M. Scibetta, R. Chaouadi, and M. W eber
SCK-CEN, Mol, Belgium
УДК 539.4
Прочность сварных соединений корпусов реакторов ВВЭР-440 в
условиях радиационного охрупчивания, отжига и повторного
охрупчивания
E. Лукон, E. ван Валле, М. Скибетта, Р. Ш оуди, М. Вебер
SCK-CEN, Мол, Бельгия
Исследованы механические свойства материалов сварных соединений корпусов реакторов
ВВЭР-440 с учетом различных состояний: исходного (неохрупченного), после радиационного
охрупчивания со средней плотностью потока нейтронов до 6,5 ■ 1019 нейтр/см2, охрупченного
с последующим отжигом, повторно охрупченного с флюэнсом частиц 1,3 ■ 1020 нейтр/см2.
Для каждого состояния материала проведены испытания на растяжение, ударное разру
шение и трещиностойкость с использованием образцов Шарпи (стандартные, восстанов
ленные, с предварительно нанесенной трещиной).
Ключевые слова: сварное соединение, радиационное охрупчивание, отжиг,
повторное охрупчивание, образец.
Introduction. The objective of SCK-CEN in contributing to the IAEA
Round-Robin exercise on the WWER-440 weld material was twofold, namely:
1. gaining experience in the field of WWER-440 pressure vessel steels;
2. analyzing Round Robin (R-R) data according to models used and
developed at SCK-CEN, in order to check their validity and applicability to this
class of steels.
In this paper we present an overview of the activity performed in SCK-CEN
during the whole exercise, from 1998 to 2001. More detailed information can be
found in [1-3].
M aterial and Test M atrix. SCK-CEN purchased part 2 of weld No. 502,
representative for WWER-440 weld material and identified as block 502.2. The
certification report includes information on weld manufacturing, heat treatment,
chemical composition and mechanical test results [4]. The cutting scheme of the
502.2-block was made according to the procedure stipulated in the terms of
reference, IAEA TC Project RER/9/035 WWER-SC-192 [5].
The specimens extracted were of Charpy-V impact type (Cv), precracked
Charpy-V (PCCv), tensile (T) and sub-size Charpy-V (MCv, 3 X 4 X27 mm); they
© E. LUCON, E. van WALLE, M. SCIBETTA, R. CHAOUADI, M. WEBER, 2004
30 ISSN 0556-171X. Проблемы прочности, 2004, № 1
SCK-CEN Contribution to the IAEA Round Robin Exercise
were subdivided into four groups, each subjected to the same mechanical tests
(tensile, impact on standard and sub-size Charpy-V specimens, fracture toughness
on PCCv):
1. unirradiated (baseline, BL);
2. irradiated (I) in the CHIVAS-7 campaign up to an average neutron fluence
of 6.5 -1019 n/cm2 (E > 1 MeV);
3. irradiated in CHIVAS-7 and then annealed (IA) as specified in the terms
of reference [100 hours at (475 ± 10)oC];
4. irradiated in CHIVAS-7, annealed and finally re-irradiated (IAR) in the
20 2
CHIVAS-8 campaign up to an average cumulated fluence of 1.3-10 n/cm
(E > 1 MeV).
Irradiations. The irradiations were carried out in the BR2 Material Testing
reactor of SCK-CEN in Mol (Belgium). Two campaigns with acronyms
CHIVAS-7 and CHIVAS-8 were carried out in the CALLISTO (CApabiLity for
Light water Irradiation in Steady state and Transient Operation) loop under PWR
conditions. Three separate legs within the loop, so-called in-pile sections (IPS),
allow to irradiate up to 162 Charpy-type specimens at the same time.
CHIVAS-7. The first irradiation, which involved 183 specimens, was
conducted using two legs (IPS1 and IPS3) of the CALLISTO loop at a
temperature of2680C and a pressure of approximately 140 bar.
The baskets were rotated in the middle of the irradiation period in order to
achieve a uniform fluence on all the specimens; the temperature of the capsule
was recorded and found to be (268^2 )0C. All specimens were in direct contact
with circulating water.
The capsule was instrumented with extended dosimetry: three dosimeters,
uniquely identified, placed at three levels in 5 of the 9 rods per each leg.
CHIVAS-8 . The second irradiation, which involved 78 specimens (60 of
which of WWER-440 weld material and the remaining of a different steel), was
conducted using one leg (IPS1) of the CALLISTO loop at a temperature of 2680C
and a pressure of approximately 155 bar.
The baskets were rotated in the middle of the 24-day irradiation cycle in
order to achieve a uniform fluence on all specimens; the temperature of the
capsule was recorded and found to be (268^2 )0C. All specimens were in direct
contact with circulating water.
As in the case of CHIVAS-7, the capsule was instrumented with extended
dosimetry: three dosimeters, uniquely identified, were placed at three levels in 5
of the 9 rods used.
Evaluation o f Specimen Fluence. The values of fluence given by the
dosimeters placed in every rod were fitted with second-order polynomials, in
order to evaluate the fast neutron fluence corresponding to the midplane of every
specimen irradiated in the two campaigns.
For rods which were not instrumented with dosimeters, the fluence was
calculated using fitting coefficients obtained from the mean of the three adjacent
instrumented rods.
The mean fluences and relevant standard deviations reported in Table 1 were
calculated for the individual specimen groups.
ISSN 0556-171X. npoôëeubi npounocmu, 2004, № 1 31
E. Lucon, E. van Walle, M. Scibetta, et al.
T a b l e 1
Mean Values and Standard Deviations Calculated for the Fluence
of the Different Specimen Groups
Condition Irradiation Specimen type Mean fluence
(n/cm2)
Standard deviation
(%)
I CHIVAS-7 T 6.5-1019 1.8
Cv 6.6-1019 5.6
MCv 6.9-1019 1.3
PCCv 6.6-1019 1.5
IA CHIVAS-7 T 6.6-1019 2.6
Cv 6.9-1019 2.4
MCv 6.9-1019 1.3
PCCv 6.7-1019 2.1
IAR CHIVAS-7
+
CHIVAS-8
T 1.3 -1020 2.2
Cv 1.3 -1020 2.1
MCv 1.3-1020 2.1
PCCv 1.3 -1020 2.2
Note: IAR correspond to accumulated fluence of two irradiations.
It was observed that the maximum difference in fluence among specimen
groups does not exceed 6%,* whereas considering individual specimens, due to
the flux curvature, it can reach almost 25%.
H eat Treatm ent Perform ed on IA and IA R Specimens. An annealing heat
treatment was performed on specimens from the IA and IAR groups.
This consisted in 100 h at (475!^ )°C, with a heating rate of 19°C/h and a
cooling rate of 30°C/h (in the oven).
This annealing treatment was proven in the past to be the most effective in
terms of Charpy-V transition temperature recovery for WWER-440 pressure
vessel steels: the lower boundary ( -2 a ) of the recovery percentage is about 80%
[6, 7].
Tensile Testing. Tensile tests have been performed in all material
conditions, in accordance with ASTM E8M on sub-size cylindrical specimens
with diameter D = 3 mm and reduced section length L0 = 14 mm (nominal
dimensions). No extensometer was used to record specimen elongation.
All tensile strength values have been fitted using the following function:
nT
a = a ath + a 0 e , (1)
where a ath represents the athermal component, while a 0 and n are fitting
parameters.
* The consequences of such variations can be considered small: according to the Russian
legislation, at such fluence levels (6.5 • 1019 n/cm2) variations of ± 10% produce a change of ± 6°C
in the (predicted) transition temperature shift.
32 ISSN 0556-171X. npoôëeMbi npounocmu, 2004, № 1
SCK-CEN Contribution to the IAEA Round Robin Exercise
Yield stress values measured in the four different conditions (BL, I, IA, IAR)
are compared in Fig. 1. Specific tensile test results can be found in [1-3].
We observed that:
(i) the annealing treatment produces a complete recovery of tensile
properties with respect to the unirradiated condition;
(ii) tensile properties measured after irradiation and re-irradiation are
equivalent, but for condition I are somewhat higher (hardening is slightly more
pronounced);
(iii) trends are generally consistent both for the yield stress and the ultimate
stress values.
Temperature (°C)
Fig. 1. Comparison of yield stress values measured in different conditions.
S tandard Charpy-V Im pact Testing. Impact tests have been performed in
all material conditions, using a calibrated 300 J pendulum equipped with an
instrumented DIN striker and a conditioning chamber which guaranteed a
temperature stability of ± 2°C. All tests have been carried out in accordance with
ASTM E23 (striker geometry excluded) on full-size (1 0x10X55 mm) Charpy-V
specimens. Specific data can be found in [1-3].
Index temperatures calculated from the impact transition curves obtained
from the various specimen groups, including baseline, are compared in Table 2;
these temperatures correspond to predefined values of dial energy (41 J, 68 J),
SFA (50%) and lateral expansion (0.89 mm), plus transition temperatures TK as
defined in the Russian regulatory code [8]. Temperature shifts are normally used
to quantify the variation of mechanical properties induced by neutron exposure or
heat treatments (e.g., annealing).
Overall results in terms of dial energy are shown in Fig. 2.
We observed that:
♦ the shift of transition curves due to irradiation is higher than the shift
caused by re-irradiation, while the relative decrease of upper shelf is more
pronounced for re-irradiation (34%) than for irradiation (26%);
ISSN 0556-171X. npoôëeubi npounocmu, 2004, № 1 33
E. Lucon, E. van Walle, M. Scibetta, et al.
♦ the annealing treatment substantially recovers the impact properties of the
baseline condition in the transition region, while the upper shelf energy is
significantly increased (31%);
♦ the almost complete recovery in terms of transition temperature is in
agreement with literature data on WWER-440 type steels (base and weld) for the
same annealing conditions [9, 10];
♦ the embrittlement caused by re-irradiation is lower than for the irradiation,
both in the transition region and in the upper shelf regime;
♦ the shifts of the transition temperature TK, calculated according to the
Russian regulatory code, are in agreement with those of the “conventional” index
temperatures.
T a b l e 2
Index Temperatures Measured on Full-Size Charpy Specimens in the Different Conditions
Investigated, and Relevant Shifts Caused by Irradiation and/or Annealing
Condition T4 J °C W °C USE, J T0.89mm, C FATT50, °C T ,
BL -6.5 13.0 119.1 -3.4 12.8 -8
I 111.7 148.9 88.5 123.4 115.2 124
IA -12.2 14.9 156.1 -0.7 23.1 -7
IAR 83.6 120.4 102.4 92.4 100.5 95
Shift AT41J, °C at68j, 0c AUSE, J AT0.89mm, C afatt50, 0c ATK, oc
BL ^ I 118.2 135.9 -30.6 126.8 102.4 132
I ^ IA -123.9 -134.0 67.6 -124.1 -92.1 -131
BL ^ IA -5.7 1.9 37.0 2.7 10.3 -1
IA ^ IAR 95.8 105.5 -53.7 93.1 70.4 102
I ^ IAR -28.1 -28.5 13.9 -31.0 -21.7 -29
Fig. 2. Comparison of dial energy impact transition curves measured on full-size Charpy-V
specimens in different conditions.
34 ISSN 0556-171X. npoôëeuu npouHocmu, 2004, № 1
SCK-CEN Contribution to the IAEA Round Robin Exercise
Sub-Size Charpy-V Im pact Testing. Sub-size impact tests have been
performed in all material conditions, using a WOLPERT PW50 pendulum
equipped with an instrumented DIN striker and a conditioning chamber which
guaranteed a temperature stability of ± 2°C. Test temperatures have been adjusted
to account for the transfer time from the environmental chamber to the impact
position, as described in [2, 11]. Tests have been carried out on sub-size
(3 X 4 x 27 mm) Charpy-V specimens; specific data can be found in [1-3].
Index temperatures calculated from the impact transition curves obtained
from the various sub-size specimen groups, including baseline, are compared in
Table 3 with relevant shifts; such temperatures correspond to predefined values of
dial energy (1.9 J, 3.1 J), SFA (50%) and lateral expansion (0.3 mm). The 1.9 J,
3.1 J and 0.3 mm levels for sub-size specimens correspond, respectively, to the
41 J, 68 J and 0.89 mm levels for full-size samples.
Overall results in terms of dial energy are shown in Fig. 3.
T a b l e 3
Index Temperatures Measured on Sub-Size Charpy Specimens in the Different Conditions
Investigated, and Relevant Shifts Caused by Irradiation or Annealing
Condition 71.9j, °C T3.1 j, °C USE, J T0.3 mm, °C FATT50, °C
BL -57.2 -46.7 6.0 -57.3 - 42.9
I 7.8 29.7 4.4 13.9 16.9
IA -59.3 -43.8 7.2 -51.1 - 47.5
IAR 0.3 19.5 4.6 0.5 14.6
Shift ArL9j, °C AT3.J °C AUSE, J AT0.3 mm, °C afatt50, 0c
BL ^ I 65.0 76.4 -1.6 71.2 59.8
I ^ IA -67.1 -73.5 2.8 -65.0 - 64.4
BL ^ IA -2.1 2.9 1.2 6.2 -4.6
IA ^ IAR 59.6 63.3 -2.6 51.6 62.1
I ^ IAR -7.5 -10.2 0.2 13.4 -2.3
--------------|A
О
---------------------------------------- BL
□
i l 111111 i9 i 111111 h* 111111 A R
a .............Д-...........
I (Chivas 7)
I + Annealed (Chivas 7)
+ Re-irr (Chivas 8)
-150 -100 -50 0 50 100 150 200 250 300
Tem perature [ X ]
Fig. 3. Comparison of dial energy impact transition curves measured on sub-size Charpy-V
specimens in different conditions.
ISSN 0556-171X. Проблемы прочности, 2004, № 1 35
E. Lucon, E. van Walle, M. Scibetta, et al.
We observed that:
the shift of the transition curves due to irradiation is almost the same as that
caused by re-irradiation, while the decrease of upper shelf is more pronounced for
re-irradiation (33%) than for irradiation (27%);
the annealing treatment substantially recovers the toughness properties of the
baseline condition in the transition region, with a significant improvement in the
upper shelf energy (15%), which is not observed for lateral expansion values;
the two irradiated conditions (I and IAR) appear to be almost identical,
particularly in terms of SFA.
The first two observations confirm the data already obtained for full-size
specimens.
Index Temperature Shifts Caused by Irradiation/Annealing M easured with
Full-Size and Sub-Size Charpy-VSpecimens. The variation of impact properties
(index temperatures and USE) can be evaluated for the four conditions under
examination considering both full-size and sub-size Charpy-V specimens.
In absolute terms, index temperature shifts measured with sub-size
specimens appear consistently lower than those yielded by full-size samples. This
trend appears specific of the WWER-440 steel, since this trend has not been
observed on other pressure vessel steels, characterized at SCK-CEN using
full-size and sub-size Charpy specimens [11] (Fig. 4).
Fig. 4. Comparison of the absolute index temperature shifts measured using full-size and sub-size
Charpy-V specimens, for the WWER-440 steel and with respect to other RPVS.
F ractu re Toughness Testing. Fracture toughness tests have been performed
in all material conditions, using a certified INSTRON universal machine and a
PID-controlled chamber which guarantees a temperature stability of ± 0.5°C
measured on the specimen.
36 ISSN 0556-171X. npoôëeubi npounocmu, 2004, № 1
SCK-CEN Contribution to the IAEA Round Robin Exercise
Tests have been carried out on fatigue precracked and 20% side-grooved
Charpy-V specimens, instrumented with clip-gage for the measurement of CMOD
(Crack Mouth Opening Displacement). Toughness data have been analyzed
according to the Master Curve methodology given in the ASTM E1921-97
standard, in order to establish the reference temperature To corresponding to a
median toughness K jc(med) = 100 MPaVm for the 1TCT reference specimen
geometry. Specific test results can be found in [2, 3].
Reference temperature values and their standard deviations are given for all
conditions in Table 4.
T a b l e 4
Reference Temperatures with Standard Deviation
Condition TO, 0C to
BL -33 5
I 66 7
IA -21 7
IAR 58 7
All test results, normalized to 1TCT, from the different conditions are
uniquely represented in Fig. 5 as a function of the difference between test
temperature and To with Master Curve and different confidence bounds (1%,
5%, 95%, 99%).
Figure 5 shows that the Master Curve methodology can effectively represent
the transitional behavior of the WWER-440 weld material in different conditions.
25
0 ^--- 1----1----1----1----1----1----1----1----1----1----1----1----1----1----1----1----1----1--- 1----1
-75 -50 -25 0 25
T - T 0 (°C)
Fig. 5. Results of the toughness tests performed in different material conditions.
If we compare temperature shifts caused by irradiation, annealing and
re-irradiation measured in terms of impact parameters (energy, SFA, lateral
expansion, and TK ) and fracture toughness (T0), we observe that T0 shifts are
ISSN 0556-171X. npoôëeMbi npounocmu, 2004, № 1 37
E. Lucon, E. van Walle, M. Scibetta, et al.
significantly smaller than those of the 41 J, 68 J, 0.89 mm or TK temperatures,
being in a good agreement (within 10°C) with the shifts of the FATT50
temperature. This circumstance was previously observed at SCK-CEN on other
pressure vessels steels, in terms of both static and dynamic fracture toughness
[12]. Furthermore, TK shifts are in better agreement with T68 J than with T41J
shifts.
Com parisons with Regulatory Lower Bound Curves. Master Curves
corresponding to low cleavage fracture probabilities (typically, 1% or 5%),
obtained in the various conditions, can be compared with toughness lower bound
curves given by various national regulatory codes. The latter represent, for reactor
pressure vessel steels, the lower limit of fracture toughness as a function of
temperature, normalized to a predefined reference transition temperature.
In this study, K Ic lower bound curves given by the Russian legislation [8]
and by the ASME code [13], widely used for Western reactor pressure vessel
steels, have been considered.
In the Russian regulatory code [8], three different static initiation toughness
curves are given, corresponding to different operating conditions: normal
conditions, upset and test conditions, and emergency conditions. The last one
(emergency conditions), corresponding to the lowest toughness level, can be used
as a lower bound curve; its expression for weld materials is the following:
On the other hand, the equation of the ASME lower bound curve (to be used
for the following steels: SA 533B Cl.1-, SA 508 Cl.2-, and SA 508 Cl.3-base
materials and welds), is
where, in the baseline condition, the Reference Temperature for Nil Ductility
Transition (RTndt) is defined as the minimum of:
♦ TNDT, measured by Pellini drop-weight tests in accordance with ASTM
E208-95a;
♦ Tcv - 33°C, where Tcv is the highest of the Charpy T68j and T0 89mm
temperatures.
In the absence of Pellini tests, the initial RTndt for the baseline condition
may be determined from Charpy tests alone; in our case: RTNDT(BL) = -20°C.
The effect of irradiation or annealing on the toughness curves is evaluated
indirectly, by shifting the curves by an amount equal to the shift of T41J from
Charpy tests; in our case:
(2)
K Ic = 36.48 + 22.783e °.°36(T -RTndt) (3)
RTNDT(I) = RTNDT(BL) + AT41J(BL^I) = 98° Q
RTNDT(IA) = RTNDT(BL) + AT41J(BL ̂IA) = - 2 6 ° C
RTNDT(IAR) = RTNDT(BL) + ^ T41J(BL^IAR) = 70°C.
38 ISSN 0556-171X. npo6n.eubi npounocmu, 2004, № 1
SCK-CEN Contribution to the IAEA Round Robin Exercise
Using ITCT-normalized fracture toughness data, the comparison between
the 1% and 5% fracture probability Master Curves and the Russian and ASME
lower bound curves is given in Fig. 6 (baseline/irradiated), Fig. 7 (irradiated/
annealed) and Fig. 8 (annealed/re-irradiated).
o I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I
-100 -75 -50 -25 0 25 50 75 100 125 150
Tem perature (°C)
Fig. 6. Comparison of the 5% and 1% fracture probability Master Curves and regulatory lower
bound curves for the baseline and irradiated conditions.
0 I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I
-100 -75 -50 -25 0 25 50 75 100 125 150
Tem perature (°C)
Fig. 7. Comparison of the 5% and 1% fracture probability Master Curves and regulatory lower
bound curves for the irradiated and annealed conditions.
In different conditions, only the ASME curve and 1% Master Curve provide
an effective lower bound to all the toughness values measured at the respective
test temperatures.
The Master Curve corresponding to 1% fracture probability results are too
conservative in the BL and IA condition; while in condition I, the 5% fracture
probability Master Curve is strongly non-conservative.
ISSN 0556-171X. npoôëeubi npounocmu, 2004, № 1 39
E. Lucon, E. van Walle, M. Scibetta, et al.
ASME
/
/ I Russian
5% CB} ~
1 % CB
0 I i i i i I i i i i I i i i i I i i i i I i i i i I i i i i I i i i i I i i i i I i i i i I i i i i
-100 -75 -50 -25 0 25 50 75 100 125 150
Temperature (°C)
Fig. 8. Comparison of the 5% and 1% fracture probability Master Curves and regulatory lower
bound curves for the annealed and re-irradiated conditions.
Evaluation of M aterial Behavior under Irradiation.
Transition Temperature Shifts. Various correlations are available for
predicting radiation-induced transition temperature shifts, based on fast neutron
fluence, chemical composition and other parameters; an extensive review is given
in [14].
The Russian correlation has been considered in this study for comparison
with temperature shifts experimentally measured for irradiation and re-irradiation.
The predictive formula used to evaluate the effect of irradiation takes into
account the contents of Cu and P, as well as the fluence. The analytical expression
is
following equivalence has been proposed [14]:
^ in 1022 n/m 2, E > 0.5 MeV ^ 18^ in 1019 n/cm2, E > 1 MeV,
which finally leads to
(4)
22 2where ^ is the neutron fluence (measured in 10 n/m , E > 0.5 MeV) and in
the case of welds (Tirr = 270oC):
Af = 800(P + 0.07Cu). (5)
19 2Since the fuence is normally given in units of 10 n/cm , E >1 MeV, the
(6)
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SCK-CEN Contribution to the IAEA Round Robin Exercise
This formula is supposed to provide an upper bound to the transition temperature
shift.
The calculated shifts of transition temperatures, both in terms of impact (TK ,
T41j) and toughness (T0) properties, measured as a consequence of irradiation
(CHIVAS-7) and re-irradiation after annealing (CHIVAS-8), are compared in
Fig. 9 with the prediction given by the above formula. Although not fully
relevant, the results obtained with the US Regulatory Code [15] and NUREG
refined correlation [16] are also plotted in Fig. 9.
For the CHIVAS-8 campaign, the reference condition is, in this case, the
annealed (IA) condition, while only the fuence accumulated in the second
irradiation (and not the total accumulated fuence) is taken into account.
Figure 9 shows that the experimentally measured shifts are underestimated
by the E900 correlation and overestimated by the ‘conservative’ Russian and
NUREG formulas.
250
P 200
e
ViÜJ
= 150 TO c
Q.
E
a io o
— o
55
E 50I—
0
5 5.5 6 6.5 7 7.5 8
Fast neutron fluence (1019 nfcm 2, E > 1 MeV)
Fig. 9. Comparison of measured and predicted transition temperature shifts.
Upper S h e lf Energy Degradation. In the Russian and US regulatory codes,
no predictive formulas are given for evaluating the decrease in USE as a
consequence of neutron irradiation.
However, in the previously mentioned NUREG report [16], a correlation is
proposed for the prediction of the Charpy Upper Shelf Energy of the irradiated
material (USEt), as a function of:
♦ USE of the unirradiated condition (USEu);
♦ product form;
♦ Cu, Ni and P contents;
♦ fast neutron fluence (0).
The correlation has the form
/ , \ 0.2223
USEt = — 0.057USEU'456 + T - [17 .5 /(Cu)(1 + 1.17Ni0 8894 ) + 3 0 5 P ] ^ 9 J (7)
ISSN 0556-171X. npoôëeubi npounocmu, 2004, № 1 41
E. Lucon, E. van Walle, M. Scibetta, et al.
with
1 1 fCu - 0.138
f (Cu) = - + - tanh I--------------
J 2 2 | 0.0846
(8)
All USE values are in ft-lb and the constant r is 55.4 for welds, 61.0 for
plates and 66.3 for forgings. The reported standard error of the model is 15 J.
For the material under examination, the results given in Table 5 have been
obtained. Once again, for the IAR condition the USE of the IA state was taken as
a reference.
T a b l e 5
Comparison between Predicted and Measured USE Values for the I and IAR Conditions
Condition USEfred, J USEmeas, J A USEpred-meas j
I 85.0 88.5 -3.5
IAR 110.0 102.4 7.6
In both cases, predictions are in close agreement with measured values. In
principle this is coincidental, as the NUREG formula is not ‘based’ on Russian
steel. However, as the transition temperature shift predictions between Russian
and NUREG codes also coincide, the correspondence might have a more
fundamental explanation.
Recovery of M aterial Toughness Due to Annealing. The key elements
with respect to continued operation of a reactor vessel after annealing are the
degree of recovery and the re-embrittlement trend.
For WWER-440 reactor pressure vessel steels, 475°C is nowadays used as
annealing temperature. Investigations done with base and weld materials,
irradiated both in power and test reactors, have shown that the difference between
annealing (Tann) and irradiation (Tirr) temperatures correlates well with the
percentage of recovery (^), while differences in the chemical composition seem to
play a minor role [9, 10].
The following empirical correlations have been reported for WWER-440
RPV steels:
v = 0.494(Tann - T irr) + 2.44, (9)
^ = 0.472(Tann - T irr ) + 5.8. (10)
In this study, (Tann - Tirr) = 207°C. Both formulas give ^ > 100%, which agrees
well with the measured recovery in terms of impact properties (TK and T41j,
^ = 99-100%). The recovery in terms of fracture toughness (T0) is however
slightly inferior (^ = 88%).
In terms of Upper Shelf Energy recovery, recent studies [17] have shown that
USE often over-recovers as a consequence of annealing; this may also explain
why the USE value after re-irradiation might not decrease significantly with
respect to the unirradiated condition.
42 ISSN 0556-171X. npoôëeubi npounocmu, 2004, № 1
SCK-CEN Contribution to the IAEA Round Robin Exercise
These assumptions are in good agreement with the results presented in this
study, since the degree of USE recovery after annealing corresponds to 131% and
the USE after re-irradiation is 86% of the unirradiated (baseline) value.
Evaluation of R e-Irrad iation Em brittlem ent. Data from studies conducted
on Western and Eastern (WWER-type) pressure vessel steels suggest that the
re-embrittlement rate (consequent to re-irradiation after annealing) is the same or
lower than in the original irradiation [18-21].
Since the available data base of annealing recovery and re-embrittlement
trend does not cover all material and annealing conditions, several assumptions
have been made in developing a trend curve approach.
The ASTM E509-97 standard, for instance, proposes two methodologies for
developing a post-annealing re-irradiation trend curve:
1. The “lateral shift methodology”: the initial irradiation (I) embrittlement
path (given by one of the predictive formulas previously discussed) is shifted
laterally at the ordinate corresponding to the annealed (IA) condition, in order to
become the post-anneal re-irradiation trend curve for the irradiated-annealed-
re-irradiated (IAR) condition. This approach postulates that re-irradiation
embrittlement follows the same rate as in the initial embrittlement stage at the
same horizontal level.
2. The “vertical shift methodology”: a portion of the initial irradiation trend,
projected as re-irradiation behaviour, is translated down vertically at the abscissa
corresponding to the irradiated (I) and annealed (IA) conditions. This way, the
re-embrittlement rate is supposed to follow the original embrittlement path as if
annealing had no influence.
In addition, a third methodology has been proposed and applied in Russia
[18, 22]:
3. The “conservative shift methodology”: the re-irradiation embrittlement
rate is supposed to follow the shape of the initial embrittlement curve from a zero
fluence point (origin). If annealing recovery is complete (as in the case of the
present study), this approach becomes identical to the lateral shift method.
Fig. 10. Experimental shifts of transition temperature TK compared to theoretical predictive models
(based on the Russian legislation), represented by solid lines; dashed curves are trend-lines
connecting experimental points.
ISSN 0556-171X. npoôëeuu npouuocmu, 2004, № 1 43
E. Lucon, E. van Walle, M. Scibetta, et al.
Using the embrittlement correlation given in the Russian regulatory code,
Fig. 10 provides comparison of the experimental results with the three predictive
methodologies described above.
We observe that:
• irradiation embrittlement is inferior to that predicted by the Russian
legislation, as previously observed;
• re-irradiation embrittlement is higher than predicted by the vertical shift
method, but significantly less than according to the lateral and conservative shift
approaches;
• however, from a qualitative point of view (see dashed lines in Fig. 10), the
re-embrittlement rate appears to be lower than the embrittlement rate.
Conclusions. An overview has been given of SCK-CEN contribution to the
IAEA Round-Robin Exercise on WWER-440 RPV Weld Material Irradiation,
Annealing and Re-Embrittlement.
Specifically, the following activities have been addressed:
■ irradiation of WWER-440 specimens in the CHIVAS-7 campaign in BR2;
■ annealing heat treatment of one group of specimens;
■ re-irradiation of another group of specimens in the CHIVAS-8 campaign;
■ tensile, instrumented impact and fracture toughness tests on BL (baseline),
(irradiated), IA (irradiated+annealed) and IAR (irradiated+annealed+
re-irradiated) specimen groups.
Comparison of different specimen groups have been made in terms of
mechanical properties using different approaches (both conventional and more
refined ones).
The most significant findings can be summarized as follows.
1. Tensile tests:
• annealing produces a complete recovery of tensile strength;
• re-irradiation causes a slightly inferior hardening as compared to
irradiation.
2. Impact tests:
• the transition curve shift caused by irradiation is higher than that obtained
for re-irradiation of full-size specimens, but is almost identical to that for
sub-size specimens;
• annealing recovers the properties of the baseline condition, with a
significant increase of upper shelf energy;
• temperature shifts induced by irradiation, annealing or re-irradiation,
measured by full-size tests, are consistently underestimated by sub-size
test results.
3. Fracture toughness tests:
• although the reference temperature measured in the IA condition is higher
than for the BL condition, however the two respective values of T0 cannot
be considered different from a statistical point of view;
• similarly, the reference temperature shift caused by re-irradiation is lower
than that in the case of irradiation, but the two values of AT) are
statistically equivalent;
• the Master Curve methodology with its associated confidence bounds
gives a satisfactory representation of the fracture toughness for all
examined conditions in the transition range;
44 ISSN 0556-171X. npo6n.eubi npounocmu, 2004, № 1
SCK-CEN Contribution to the IAEA Round Robin Exercise
• ДГ0 shifts are generally lower than the corresponding shifts of the most
Charpy index temperatures, although the agreement with ДРЛТТ5о is
very good;
• Master Curves associated to low fracture probabilities (1%, 5%) have been
compared with regulatory lower bound toughness curves (Russian,
ASME).
4. In terms of neutron embrittlement, the measured transition temperature
shifts for irradiation and re-irradiation are underestimated by the present
formulation of ASTM E900-87(94) and overestimated by the Russian legislation
formula and by an improved correlation proposed by NUREG in more recent
years.
5. The complete recovery observed is in agreement with literature data for
similar materials and annealing conditions.
6. The re-embrittlement rate following re-irradiation after annealing appears
lower than the embrittlement rate from the baseline condition.
To summarize in a concise form the effects of annealing and re-irradiation
which have been observed in this work, the following can be stated:
• annealing produces a complete recovery of mechanical properties after
irradiation, with a significant increase in USE;
• the embrittlement caused by re-irradiation after annealing is inferior to the
initial toughness degradation induced by irradiation.
Acknowledgements. The authors acknowledge contribution of the technical
staff of LHMA, TCH and BR2 departments to this activity, and more specifically
to W. Claes and his team for dismantling, to R. Mertens for specimen preparation
and to L. Van Houdt, J. Veraghtert, and R. Vosch for testing.
Р е з ю м е
Досліджуються механічні характеристики зварних з ’єднань корпусів реак
торів BBEP-440 з урахуванням різного стану: початкового (неокрихчувано-
го), після радіаційного окрихчування з середньою густиною потоку нейтро-
19 2нів до 6,5-10 нейтр/см , окрихчуваного з наступним відпалом, повторно
20 2окрихчуваного з флюенсом частинок 1,3-10 нейтр/см . Для кожного
стану матеріалу проведено дослідження на розтяг, ударне руйнування та
тріщиностійкість із використанням зразків Шарпі (стандартні, відновлені та
з попередньо нанесеною тріщиною).
1. E. van Walle et al., SCK-CEN Contribution to the ІЛЕЛ Round Robin
Exercise on WWER-440 RPV Weld Material Irradiation, Annealing, and
Re-Embrittlment, SCK-CEN Report BLG-768 (1998).
2. E. van Walle et al., SCK-CEN Contribution to the IAEA Round Robin
Exercise on WWER-440 RPV Weld Material Irradiation, Annealing, and
Re-Embrittlement, 2nd Progress Report, SCK-CEN Report BLG-827 (1999).
3. E. Lucon et al., SCK-CEN Contribution to the IAEA Round Robin Exercise
on W W ER-440 R P V W eld M a ter ia l Irrad ia tion , A nnea ling , and
Re-Embrittlment, SCK-CEN Final Report BLG-845 Rev 1 (2000).
ISSN 0556-171X. Проблеми прочности, 2004, № 1 45
E. Lucon, E. van Walle, M. Scibetta, et al.
4. Investigation Results Report - Weld, No. 502 Certification, MOHT-OTJIG
RM and EDO, Gidropress (1997).
5. IAEA Round Robin Exercise on WWER-440 RPV Weld Material Irradiation,
Annealing, and Re-Embrittlement: Terms o f Reference, IAEA TC Project
RER/9/035 WWER-SC-192 (1996).
6. M. Brumovsky, P. Novosad and M. Vacek, “Material properties of
Bochunice 1 and 2 reactor pressure vessel materials before and after
annealing,” in: International Workshop on WWER-440 Reactor Pressure
Vessel Embrittlement and Annealing, Zavazna Poruba (29-31 March),
Slovak Republic (1994).
7. R. Pelli and K. Torronen, State o f the Art Review on Thermal Annealing,
AMES Report No. 2, EUR 16278 EN (1995).
8. PNAE-G-7-002-86. Calculation Standard fo r Strength o f Equipment and
Pipes o f Nuclear Power Units [in Russian], Energoatomizdat, NGA-01-85-1,
NIKIET (1989).
9. A. D. Amaev et al., “Study of mechanical properties recovery processes of
irradiated pressure vessel steels at annealing,” in: Radiation Materials
Science [in Russian], Vol. 4, Kharkov, pp. 25-32.
10. V. A. Ignatov et al., “Annealing of irradiation embrittlement of reactor
pressure vessel material in Novovoronezh 1 NPP,” in: Joint Seminar on
Irradiation Embrittlement o f Reactor Pressure Vessels (Cologne, Sep. 3-7)
(1990).
11. E. Lucon et al., Correlations between Standard and Miniaturized Charpy-V
Specimens, SCK-CEN Report BLG-797 (1998).
12. E. Lucon and R. Chaouadi, “Radiation damage assessment by the use of
dynamic toughness measurements on pre-cracked Charpy-V specimens”, in:
Effects o f Radiation on Materials (20th Int. Symp.); S. T. Rosinski, M. L.
Grossbeck, T. R. Allen, and A. S. Kumar (Eds.), ASTM STP 1405, West
Conshohocken, PA (2002).
13. ASME Boiler and Pressure Vessel Code, Section II (1995).
14. P. Petrequin, A Review o f Formulas fo r Predicting Irradiation Embrittlement
o f Reactors Vessel Material, AMES Report No. 5, EUR 16455 EN (1996).
15. P. N. Randall, “Basis for revision 2 of U.S. NRC Regulatory Guide 1.99,” in:
Radiation Embrittlement o f Nuclear Reactor Pressure Vessel Steels, ASTM
STP 909 (1986), pp. 149-162.
16. E. D. Eason, J. E. Wright, and G. R. Odette, Improved Embrittlement
Correlations fo r Reactor Pressure Vessel Steels, NUREG/CR-6551, MCS
970501 (1998).
17. S. K. Iskander et al., “Re-irradiation response rate of a high-copper reactor
pressure vessel weld,” presented at the 20th ASTM Symposium on the Effect
o f Radiation on Materials (Williamsburg Virginia, June 6-8) (2000).
18. A. D. Amaev, A. M. Kryukov, and M. A. Sokolov, “Recovery of the
transition temperature of irradiated WWER-440 vessel material by annealing,”
in: Radiation Embrittlement o f Nuclear Reactor Pressure Vessel Steels,
ASTM STP 1170, 1193 (1992), pp. 369-379.
46 ISSN 0556-171X. npo6n.eubi npounocmu, 2004, № 1
SCK-CEN Contribution to the IAEA Round Robin Exercise
19. T. R. Mager, “Thermal annealing of an embrittled reactor vessel: feasibility
and methodology,” in: Nuclear Engineering and Design, Vol. 124 (1990),
pp. 43-51.
20. J. R. Hawthorne, H. E. Watson, and F. J. Loss, Experimental investigation of
multi cycle irradiation and annealing effects on notch ductility of A533-B
weld deposits, in: ASTM STP 725 (1981).
21. A. M. Kryukov and M. A. Sokolov, “Investigation of material behavior
under rerradiation after annealing using subsize specimens,” in: Small
Specimen Test Techniques Applied to Nuclear Reactor Vessel Thermal
Annealing and Plant Life Extension, ASTM STP 1204 (1993), pp. 417-423.
22. Reactor Pressure Vessel Embrittlement. Report o f the IAEA Extrabudgetary
Programme on the Safety o f WWER-440 Model 230 Nuclear Power Plants,
IAEA-TECDOC-659 (1992).
Received 26. 05. 2003
ISSN 0556-171X. npoôëeuu npouuocmu, 2004, № 1 47
|
| id | nasplib_isofts_kiev_ua-123456789-47059 |
| institution | Digital Library of Periodicals of National Academy of Sciences of Ukraine |
| issn | 0556-171X |
| language | English |
| last_indexed | 2025-12-07T19:03:15Z |
| publishDate | 2004 |
| publisher | Інститут проблем міцності ім. Г.С. Писаренко НАН України |
| record_format | dspace |
| spelling | Lucon, E. Walle, E. van Scibetta, M. Chaouadi, R. Weber, M. 2013-07-09T13:30:16Z 2013-07-09T13:30:16Z 2004 SCK-CEN Contribution to the IAEA Round Robin Exercise on
 WWER-440 RPV Weld Material: Irradiation, Annealing, and
 Re-Embrittlement / E. Lucon, E. van Walle, M. Scibetta, R. Chaouadi, M. Weber // Проблемы прочности. — 2004. — № 1. — С. 30-47. — Бібліогр.: 22 назв. — англ. 0556-171X https://nasplib.isofts.kiev.ua/handle/123456789/47059 539.4 Mechanical properties of WWER-440 RPV weld joints have been studied with account of different states of the material: baseline (unirradiated), irradiated up to the average fast neutron fluence of 6.5*10^19 n/cm^2, irradiated and eventually annealed, re-irradiated with the accumulated fast neutron fleunce 1.3*10^20 n/cm^2. For each state of the material, tensile, impact fracture, and fracture toughness tests were performed with the use of Charpy (standard, reconstituted, and pre-cracked) specimens. Исследованы механические свойства материалов сварных соединений корпусов реакторов ВВЭР-440 с учетом различных состояний: исходного (неохрупченного), после радиационного охрупчивания со средней плотностью потока нейтронов до 6,5*10^19 нейтр/см^2, охрупченного с последующим отжигом, повторно охрупченного с флюэнсом частиц 1,3*10^20 нейтр/см^2. Для каждого состояния материала проведены испытания на растяжение, ударное разрушение и трещиностойкость с использованием образцов Шарпи (стандартные, восстановленные, с предварительно нанесенной трещиной). Досліджуються механічні характеристики зварних з ’єднань корпусів реакторів BBEP-440 з урахуванням різного стану: початкового (неокрихчувано- го), після радіаційного окрихчування з середньою густиною потоку нейтронів до 6,5*10^19 нейтр/см^2, окрихчуваного з наступним відпалом, повторно окрихчуваного з флюенсом частинок 1,3*10^20 нейтр/см^2. Для кожного стану матеріалу проведено дослідження на розтяг, ударне руйнування та тріщиностійкість із використанням зразків Шарпі (стандартні, відновлені та з попередньо нанесеною тріщиною). en Інститут проблем міцності ім. Г.С. Писаренко НАН України Проблемы прочности Научно-технический раздел SCK-CEN Contribution to the IAEA Round Robin Exercise on WWER-440 RPV Weld Material: Irradiation, Annealing, and Re-Embrittlement Прочность сварных соединений корпусов реакторов ВВЭР-440 в условиях радиационного охрупчивания, отжига и повторного охрупчивания Article published earlier |
| spellingShingle | SCK-CEN Contribution to the IAEA Round Robin Exercise on WWER-440 RPV Weld Material: Irradiation, Annealing, and Re-Embrittlement Lucon, E. Walle, E. van Scibetta, M. Chaouadi, R. Weber, M. Научно-технический раздел |
| title | SCK-CEN Contribution to the IAEA Round Robin Exercise on WWER-440 RPV Weld Material: Irradiation, Annealing, and Re-Embrittlement |
| title_alt | Прочность сварных соединений корпусов реакторов ВВЭР-440 в условиях радиационного охрупчивания, отжига и повторного охрупчивания |
| title_full | SCK-CEN Contribution to the IAEA Round Robin Exercise on WWER-440 RPV Weld Material: Irradiation, Annealing, and Re-Embrittlement |
| title_fullStr | SCK-CEN Contribution to the IAEA Round Robin Exercise on WWER-440 RPV Weld Material: Irradiation, Annealing, and Re-Embrittlement |
| title_full_unstemmed | SCK-CEN Contribution to the IAEA Round Robin Exercise on WWER-440 RPV Weld Material: Irradiation, Annealing, and Re-Embrittlement |
| title_short | SCK-CEN Contribution to the IAEA Round Robin Exercise on WWER-440 RPV Weld Material: Irradiation, Annealing, and Re-Embrittlement |
| title_sort | sck-cen contribution to the iaea round robin exercise on wwer-440 rpv weld material: irradiation, annealing, and re-embrittlement |
| topic | Научно-технический раздел |
| topic_facet | Научно-технический раздел |
| url | https://nasplib.isofts.kiev.ua/handle/123456789/47059 |
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