Radiation-Induced Embrittlement of WWER-440 Reactor Pressure Vessel Steel under Loading
Приведены результаты определения эталонной температуры T0 и построена "Master curve” на основе экспериментов, выполненных для стали марки 15Х2МФА (основной металл корпуса реактора типа ВВЭР-440) в трех состояниях: необлученном, облученном и облученном под нагрузкой. Показано, что механическая н...
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Інститут проблем міцності ім. Г.С. Писаренко НАН України
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| Zitieren: | Radiation-Induced Embrittlement of WWER-440 Reactor Pressure Vessel Steel under Loading / E.U. Grinik , L.I. Chirko , Yu.S. Gul’chuk, V.A. Strizhalo, L.S. Novogrudskii, A. Ballesteros, L. Debarberis, F. Sevini // Проблемы прочности. — 2003. — № 1. — С. 39-47. — Бібліогр.: 10 назв. — англ. |
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irk-123456789-469402013-07-08T08:46:25Z Radiation-Induced Embrittlement of WWER-440 Reactor Pressure Vessel Steel under Loading Grinik, E.U. Chirko, L.I. Gul’chuk, Yu.S. Strizhalo, V. A. Novogrudskii, L.S. Ballesteros, A. Debarberis, L. Sevini, F. Научно-технический раздел Приведены результаты определения эталонной температуры T0 и построена "Master curve” на основе экспериментов, выполненных для стали марки 15Х2МФА (основной металл корпуса реактора типа ВВЭР-440) в трех состояниях: необлученном, облученном и облученном под нагрузкой. Показано, что механическая нагрузка, имитирующая давление теплоносителя, ускоряет радиационное охрупчивание, причем вклад ее сравним с вкладом нейтронного облучения. Наведено результати визначення еталонної температури T0 та побудовано “Master curve” на основі експериментів, проведених для сталі марки 15Х2МФА (основний метал корпусу реактора типу ВВЕР-440) в неопроміненому, опроміненому й опроміненому під навантаженням стані. Показано, що механічне навантаження, яке імітує тиск теплоносія, прискорює радіаційне окрих- чення, при цьому його внесок можна порівняти з внеском нейтронного опромінення. 2003 Article Radiation-Induced Embrittlement of WWER-440 Reactor Pressure Vessel Steel under Loading / E.U. Grinik , L.I. Chirko , Yu.S. Gul’chuk, V.A. Strizhalo, L.S. Novogrudskii, A. Ballesteros, L. Debarberis, F. Sevini // Проблемы прочности. — 2003. — № 1. — С. 39-47. — Бібліогр.: 10 назв. — англ. 0556-171X http://dspace.nbuv.gov.ua/handle/123456789/46940 539.4 en Проблемы прочности Інститут проблем міцності ім. Г.С. Писаренко НАН України |
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Научно-технический раздел Научно-технический раздел |
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Научно-технический раздел Научно-технический раздел Grinik, E.U. Chirko, L.I. Gul’chuk, Yu.S. Strizhalo, V. A. Novogrudskii, L.S. Ballesteros, A. Debarberis, L. Sevini, F. Radiation-Induced Embrittlement of WWER-440 Reactor Pressure Vessel Steel under Loading Проблемы прочности |
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Приведены результаты определения эталонной температуры T0 и построена "Master curve” на основе экспериментов, выполненных для стали марки 15Х2МФА (основной металл корпуса реактора типа ВВЭР-440) в трех состояниях: необлученном, облученном и облученном под нагрузкой. Показано, что механическая нагрузка, имитирующая давление теплоносителя, ускоряет радиационное охрупчивание, причем вклад ее сравним с вкладом нейтронного облучения. |
| format |
Article |
| author |
Grinik, E.U. Chirko, L.I. Gul’chuk, Yu.S. Strizhalo, V. A. Novogrudskii, L.S. Ballesteros, A. Debarberis, L. Sevini, F. |
| author_facet |
Grinik, E.U. Chirko, L.I. Gul’chuk, Yu.S. Strizhalo, V. A. Novogrudskii, L.S. Ballesteros, A. Debarberis, L. Sevini, F. |
| author_sort |
Grinik, E.U. |
| title |
Radiation-Induced Embrittlement of WWER-440 Reactor Pressure Vessel Steel under Loading |
| title_short |
Radiation-Induced Embrittlement of WWER-440 Reactor Pressure Vessel Steel under Loading |
| title_full |
Radiation-Induced Embrittlement of WWER-440 Reactor Pressure Vessel Steel under Loading |
| title_fullStr |
Radiation-Induced Embrittlement of WWER-440 Reactor Pressure Vessel Steel under Loading |
| title_full_unstemmed |
Radiation-Induced Embrittlement of WWER-440 Reactor Pressure Vessel Steel under Loading |
| title_sort |
radiation-induced embrittlement of wwer-440 reactor pressure vessel steel under loading |
| publisher |
Інститут проблем міцності ім. Г.С. Писаренко НАН України |
| publishDate |
2003 |
| topic_facet |
Научно-технический раздел |
| url |
http://dspace.nbuv.gov.ua/handle/123456789/46940 |
| citation_txt |
Radiation-Induced Embrittlement of WWER-440 Reactor Pressure
Vessel Steel under Loading / E.U. Grinik , L.I. Chirko , Yu.S. Gul’chuk, V.A. Strizhalo, L.S.
Novogrudskii, A. Ballesteros, L. Debarberis, F. Sevini // Проблемы прочности. — 2003. — № 1. — С. 39-47. — Бібліогр.: 10 назв. — англ. |
| series |
Проблемы прочности |
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| fulltext |
UDC 539.4
Radiation-Induced Embrittlement of WWER-440 Reactor Pressure
Vessel Steel under Loading
E. U. G rin ik ,a L. I. C h irk o ,a Y u. S. G u l’chuk ,a V. A. S trizhalo ,b L. S.
N ovogrudskii,b A. B allesteros,c L. D eb arb eris ,d F. Sevinid
a Institute for Nuclear Researches, National Academy of Sciences of Ukraine, Kiev,
Ukraine
b Pisarenko Institute of Problems of Strength, National Academy of Sciences of Ukraine,
Kiev, Ukraine
c Tecnatom S.A., Madrid, Spain
d JRC, Institute for Energy, Petten, Netherlands
УДК 539.4
Радиационное охрупчивание корпусной стали реактора ВВЭР-440,
находящейся под нагрузкой
Э. У. Г р и н и к а, Л . И. Ч и р к о а, Ю . C. Г у л ьч у к а, В. А. С тр и ж ал о 6, Л . С.
Н овогрудски й 6, А. Б аллестеросв, Л . Д е6a р 6 ерисг, Ф. С евин и г
а Научный центр “Институт ядерных исследований” НАН Украины, Киев, Украина
6 Институт проблем прочности им. Г. С. Писаренко НАН Украины, Киев, Украина
в “Tecnatom S.A.”, Мадрид, Испания
г Энергетический институт, Петен, Нидерланды
Приведены результаты определения эталонной температуры Г0 и построена "Master
curve” на основе экспериментов, выполненных для стали марки 15Х2МФА (основной металл
корпуса реактора типа ВВЭР-440) в трех состояниях: необлученном, облученном и
облученном под нагрузкой. Показано, что механическая нагрузка, имитирующая давление
теплоносителя, ускоряет радиационное охрупчивание, причем вклад ее сравним с вкладом
нейтронного облучения.
Ключевые слова : нейтронное облучение, радиационное охрупчивание, кор
пусная сталь, “Master curve”, механическая нагрузка, давление теплоносителя.
In troduc tion . A t the time being, the radiation embrittlement o f the reactor
vessel materials is evaluated by the results o f the Charpy-type surveillance
specimens or fatigue pre-cracked compact tension specimens. The surveillance
specimens are irradiated in the hermetically closed container assemblies without
stress. The reactor vessel m aterial is actually subjected to the pressure o f the
coolant.
© E. U. GRINIK, L. I. CHIRKO, Yu. S. GUL’CHUK, V. A. STRIZHALO, L. S. NOVOGRUDSKII,
A. BALLESTEROS, L. DEBARBERIS, F. SEVINI, 2003
ISSN 0556-171X. Проблемы прочности, 2003, N2 1 39
É. U. Grinik, L. I. Chirko, Yu. S. G u l’chuk, et al.
For determining the effect o f m echanical load under irradiation on the
embrittlement rate o f the reactor vessel m aterials two container assemblies were
irradiated in commercial unit. They were completed w ith the specimens cut of
four various grades o f steel, having been utilized for manufacturing reactor
pressure vessels in the former Soviet Union. Three kinds o f specimens were
prepared o f each type o f steel, namely:
- static tensile specimens;
- Charpy-V type im pact specimens;
- T-L-oriented stressed and unstressed compact tension specimens.
The specimens o f each steel were located in three layers. There were 2
tensile, 2 Charpy-V and 4 CT (2 stressed and 2 unstressed) specimens in every
layer o f each assembly. Thus, each assembly was completed w ith six stressed and
six unstressed 1/2T compact tension specimens m anufactured o f each type of
vessel steel.
Irradiation was perform ed on such a level o f the reactor core that within one
year o f irradiation to accumulate the dose comparable w ith that accumulated by
the reactor core in the center o f the active zone during the design service life (40
years).
The prim ary material property, determining its susceptibility to spalling, is
the ability o f the material to resist the propagation o f a crack. For reactor steels
the ductile-brittle transition tem perature shift in the irradiated state has to be
defined as compared to the unirradiated one, therefore the application o f large-
size surveillance specimens is unacceptable.
Employing the M aster Curve approach for ferritic steels enables to obtain
credible values o f the critical stress-intensity factor K jc in the wide temperature
range testing small-size specimens [1]. According to the recent standardizing
documents, M aster Curve is constructed on the basis o f the results o f testing
small-size specimens at a single temperature such that steady growth o f a crack
before the beginning o f the fracture is practically precluded. This allows to
simplify considerably the procedure o f defining j -integral and use experimental
basic approach o f nonlinear fracture mechanics.
The paper presents the results o f determining the reference tem perature TO
and constructing a M aster Curve proceeding from the experiments, carried out for
the steel 15Cr2MoVA (base metal o f W W ER-440 reactor vessel metal) in
irradiated, unirradiated, and irradiated stressed states.
M ate ria l an d Specim ens. Chemical composition o f the investigated steel is
presented in Table 1. 1/2T CT specimens were studied. Fatigue cracks (a 0) on
them were initiated in line w ith the requirements o f the State Standard 25.506-85
[2 ].
T a b l e 1
Chemical Composition of 15Cr2MoVA Steel
C Si Mn Ni Cr S P Cu V Mo
0.14 0.2 0.36 0.11 2.0 0.012 0.007 0.09 0.2 0.6
The vessel metal o f an operating W W ER-type reactor is under the pressure
o f the coolant. In W W ER-1000 it equals 16 MPa, that results in the stress o f the
40 ISSN Ü556-171X. Проблемыг прочности, 2ÜÜ3, N 1
Radiation-Induced Embrittlement o f WWER-440 Reactor
vessel m etal o f about 173 MPa. For imitating such stresses m echanical load was
created on CT specimens by means o f the loading screw and m etallic bellows.
Stressed specimens were assembled into two chains, linked w ith two bellows,
next to unstressed specimens, enabling the correct comparison o f the data for both
groups o f specimens.
The design o f the assemblies allowed water o f the prim ary circuit to wash
the specimens, so their irradiation tem perature was equal to the tem perature o f the
coolant, i.e., ~ 2 9 0 oC.
Fast neutron fluences (E > 0.5 MeV) accumulated by the studied specimens
in experimental assemblies are calculated by the Kurchatov Institute in the
framework o f the Project TACIS PCP-IV [3] and are illustrated in Table 2.
T a b l e 2
Calculated Fluences Defined for Middle Areas of Each Assembly Layer
Steel grade Layer number Neutron fluence (E > 0.5 MeV), n/cm2
Assembly No. 1 Assembly No. 2
Base metal
15Cr2MoVA
7 9.15 •Ю19 9.24 •1019
8 1.04 -1020 1.05 •Ю20
9 1.16-1020 1.17 •Ю20
M ethods o f Testing. Static tensile testing was conducted in accordance with
the State Standard 1497-73 [4]. Rem ote-controlled tensile-testing machine
installed in the “hot” chamber was applied as experimental equipment. The error
o f the breaking stress determining was ± 25 MPa. The active grip m oved with the
speed o f 1 mm/min. Tests were perform ed at room temperature and at 350OC.
Keeping tem perature was w ith the accuracy ±2OC. The m aximum error of
determining strength characteristics is 5% and that o f plasticity - 2%. The results
o f static tensile tests are shown in Table 3.
T a b l e 3
Radiation-Induced Changes of Mechanical Characteristics of Base Metal 15Cr2MoVA
Averaged over Specimens of Both Assemblies
Test , OC Rp0.2,
MPa
,am
P
R
M
A>,% %%SA Z, %
20 333 441 17.0 7.3 77
601 670 5.1 2.2 44
Change, % 80 51 -70 -70 -43
350 316 387 12.3 3.6 80
449 506 5.0 2.5 43
Change, % 42 31 -59 -31 -46
Note. Over the line are given results for unirradiated specimens and under the line - for irradiated
specimens.
Fracture toughness tests were carried out in line w ith ASTM Standard
1921-97 [1] on the testing electromechanically-driven m achine Instron-8500 with
the critical load 100 kN. The m achine is remote-controlled and it is installed in
ISSN 0556-171X. Проблемыг прочности, 2003, N І 41
E. U. Grinik, L. I. Chirko, Yu. S. G u l’chuk, et al.
the “hot” chamber o f the Institute for N uclear Research o f the National Academy
o f Sciences o f Ukraine in the framework o f the program TACIS PCP-IV. The
procedure o f m easuring temperature, load and crack opening satisfied the
requirements [1] that provided determination o f the above characteristics with
high accuracy and obtaining reliable P — V diagrams.
Since crack opening was m easured on the front surface o f a specimen,
according to p. 7.1 [1] the correction for displacement o f the loaded line was
applied by m ultiplying the m easured values by 0.73.
Fracture toughness testing temperature was found as
where C = —28°C for 1/2T specimens [1]. The temperature T28j was defined by
the results o f Charpy im pact testing (1 0 X1 0 X55 mm) on the pendulum hammer
KM D-30D with im pact accumulated energy 300 J in the temperature range
(—80°C)-(+ 100°C) in line with the standard [5] (Fig. 1). The processing o f impact
toughness data (E ab) w as conducted according to [6 ] by m eans o f the
approximation function o f hyperbolic tangent, the equation o f w hich looks like
where E ab is the absorbed energy (J), A is the average value o f E ab between
the maxim um (Emax) and minimum (Emin) values, B = (E max — E min ) /2 , Tk0 is
the temperature corresponding to the value A, and C is the empiric constant.
Parameters A, B , C , and Tk0 are defined by the processing o f the experimental
points by the methods o f least squares.
Ductile-brittle transition temperature for unirradiated specimens (Tk0) is
— 62°C and for irradiated ones it is TkF = — 7°C.
T = T28 J + C , ( 1)
1— |— r— i— r— i— r 1 1 ) 1 1 l '" " T ..... | , I , M1...... 1...... 1
-100 -50 0 50 100
Test temperature (°C)
Fig. 1. The temperature dependence of absorbed energy of base metal 15Cr2MoVA.
42 ISSN 0556-171X. npoÖÄeubi npounocmu, 2003, N2 1
Radiation-Induced Embrittlement o f WWER-440 Reactor
From the data o f Fig. 1 the fracture toughness test temperature =
— 11.5°C. Thus, according to the standard [1] and on the basis o f Eq. (1) fracture
toughness tests have to be perform ed at the temperature T = — 90°C and
T = — 40°C for unirradiated and irradiated specimens, respectively.
O b ta in ed R esults. As a result o f testing seven unirradiated specimens at
— 90°C P — V diagrams, characteristic for cleavage cracking, have been got. By
the results o f processing these diagrams the values o f /- in teg ra l were defined as
the sum o f elastic and plastic components. The values o f the stress intensity factor
(K Jc) were calculated from the j c values obtained for each individual specimen.
The values o f J c and K jc for unirradiated specimens o f steel 15Cr2MoVA are
illustrated in Table 4.
P — V diagram and the fracture surface in first irradiated specimen at the
testing temperature Ttest = — 40°C testify the considerable stable growth o f a
crack. The value o f K jc for this specimen is equal to 129.95 MPaVm (Table 4).
According to the recommendations o f CSRIEM “Prom etei” on the assessment of
fracture toughness in W W ER-440, W W ER-1000 reactor vessel materials, the
testing temperature was lowered by 20°C, but at Ttest = — 60°C the stable crack
growth was also more intensive. A t Ttest = — 80°C cleavage cracking took place in
the irradiated specimens. Therefore all the rest irradiated (stressed and unstressed)
specimens were tested at Ttest = — 80°C. The results o f testing all the specimens
are shown in Table 4. None o f the K j c values should be qualified as all the
values do not exceed K jc(limit) value for the above testing temperature.
M aste r C urve A p p ro ach A pplication. M aster Curve approach is used for
construction tem perature dependence o f fracture toughness on the basis o f test
results o f lim ited quantity o f specimens [7, 8 ]. The position o f the curve K j c(T )
on the tem perature coordinate is established from experim ental determ ination o f
the tem perature T0 at w hich the m edian value K jc( med) for 1T specimens is
100 M P aV m . Ferritic steels are known to be very heterogeneous due to the
orientation o f individual grains as well as to heterogeneities o f the grain
boundaries. Carbides and different non-metallic inclusions on the grain
boundaries m ay be nucleus o f microcracks. Their random location relative to the
crack front in the m aterial conditions the large spread in values o f fracture
toughness testing small-size specimens.
As it is shown by Weibull [9] the curve shape for ferritic steels w ith yield
strength from 275 to 825 M Pa is determined by the index o f the exponential curve
b = 4 and K min = 20 MPaVm. These data are received on the basis o f the statistic
analysis o f a huge num ber o f experimental data, obtained by various researchers
all over the world on the specimens o f different sizes.
According to the requirements [1], whose adequacy has been confirmed in
[10], the values obtained on specimens o f size 0.5T were recalculated into the
values equivalent to those obtained on specimens o f size 1T by the formula
= T 35J/cm2 for unirradiated specimens was — 62°C and for irradiated ones it was
(3)
ISSN 0556-171X. npo6n.eubi npounocmu, 2003, № 1 43
E. U. Grinik, L. I. Chirko, Yu. S. G u l’chuk, et al.
T a b l e 4
Results of Fracture Toughness Testing of Base Metal 15Cr2MoVA
No. State
P3 KJc (1/2T),
MPaVm
KJc (1TH
MPaVm
K 0,
MPaVm
K ,KJc (med) •
MPaVm
0̂E-?
1 Unirradiated -90 77.57 68.42 85.58 79.83 - 72.1
2 -90 85.53 75.11
3 -90 93.65 81.94
4 -90 98.09 85.67
5 -90 98.93 86.38
6 -90 106.12 92.44
7 -90 107.09 93.24
1 Irradiated -80 67.24 59.73 81.98 76.56 - 58.5
2 unstressed -80 67.92 60.30
3 -80 73.98 64.81
4 -80 80.02 70.47
5 -80 86.28 71.84
6 -80 98.42 85.95
7 -80 103.83 88.92
8 -80 101.95 90.50
9 -80 104.93 91.43
10 -80 106.48 92.73
11 -40 129.95 112.46
12 -60 144.78 124.94
1 Irradiated -80 51.51 46.50 68.60 64.35 - 42.5
2 stressed -80 55.17 49.58
3 -80 56.45 50.66
4 -80 58.30 52.21
5 -80 67.64 60.07
6 -80 68.15 60.50
7 -80 69.26 61.42
8 -80 82.22 72.33
9 -80 82.24 72.34
10 -80 86.94 76.30
11 -80 91.83 80.41
12 -80 94.55 82.70
and then we calculated the scale parameter
K 0 =
N
( K Jc(1T))
i=1 N - 0.3068
1/4
+ 2 0 , (4)
44 ISSN 0556-171X. npo6n.eubi npounocmu, 2003, N2 1
Radiation-Induced Embrittlement o f WWER-440 Reactor
15Cr2M oVA unirrad iated
IV
I
.a
2 3 4
KJc (i) - Km
Master Curve for unirradiated 15Cr2MoVA sleel
Test temperature (0C)
a
15Cr2MoVA irradiated
IV
I
_0
£
MKJc(i) - Kmini
b
15Cr2MoVA irradiated under stress
0 1 2 3 4 5 6 7
ln[KJc (i) - Kmin]
M aster Curve lor irradiated under stress
15Cr2MoVA steel
Test temperature (0C)
c
Fig. 2. Weibull’s plots and Master Curves for three states of specimens: unirradiated (a), irradiated
without stress (b) and irradiated under mechanical load (c): (1, 3) tolerance bounds 95% and 5%,
respectively; (2) Master Curves.
and the m ean value o f the sample:
K M med) = (K 0 - 20)[ln2 ]1/4 + 20. (5)
The magnitude o f the reference tem perature (70) was determined using the
following expression:
t - T - 1 i ( Kjc(med)
0 test 0.019 n ,
30
70 (6 )
ISSN 0556-171X. npoÖÄeubi npounocmu, 2003, N 1 45
E. U. Grinik, L. I. Chirko, Yu. S. G u l’chuk, et al.
The values o f the parameters calculated according to formulas (3)-(5) for three
states o f steel 15Kh2MFA are listed in Table 4. M aster Curves (Fig. 2) were
constructed by the equation [1]:
Kjc(med) = 3 0 + 70exp[0 .019(r- r 0 )]. (7)
For visual representation o f the results o f testing, W eibull’s m odel was used
according to w hich the probability o f fracture is evaluated by the formula
P f = 1 - ex p { -[(K jc - K min ) /(K 0 - K min ) f }, (8)
where K min = 20 MPaVm and b = 4.
Figure 2 demonstrates that the data o f K Jc lie within ± 5 % confidence
interval o f M aster Curve approach.
M ean-square deviations T0 estimated according to [1] are equal ±7°C for
unirradiated specimens and ± 6°C for irradiated ones in both states.
Figure 3 illustrates M aster Curves for three states o f steel 15Cr2MoVA:
unirradiated (1), after irradiation unloaded (2), and irradiated loaded (3).
Reference temperature T0 for irradiated unstressed specimens exceeds its value
for unirradiated specimens. Total effect o f neutron irradiation and stress results in
the increase o f the reference tem perature in this case by 30°C.
Test temperature (°C)
Fig. 3. Master Curves for three states of specimens: (7) unirradiated, (2) irradiated, and (3) irradiated
under mechanical load.
C onclusions. The effect o f neutron irradiation on radiation em brittlement of
N i-free vessel steel 15Cr2MoVA is studied in two states: under m echanical load
and without it.
The effect o f stressed state due to m echanical load, im itating the pressure o f
the coolant, on the ductile-brittle transition temperature o f the fatigue pre-cracked
specimens o f steel is discovered. The above effect is revealed in the reference
temperature T0 for stressed irradiated specimens being higher than for irradiated
unstressed ones. The effect o f the stressed state caused by m echanical load,
46 ISSN 0556-171X. npo6n.eubi npounocmu, 2003, № 1
Radiation-Induced Embrittlement o f WWER-440 Reactor
imitating the pressure o f the coolant, on the ductile-brittle transition temperature
is qualitatively comparable with the effect o f neutron irradiation. From the
physical point o f view, this phenom enon is conditioned by the higher rate o f the
formation o f radiation defects due to the total effects o f neutron irradiation and
stress.
Р е з ю м е
Наведено результати визначення еталонної температури Г0 та побудовано
“Master curve” на основі експериментів, проведених для сталі марки 15Х2МФА
(основний метал корпусу реактора типу ВВЕР-440) в неопроміненому, опро
міненому й опроміненому під навантаженням стані. Показано, що механічне
навантаження, яке імітує тиск теплоносія, прискорює радіаційне окрих-
чення, при цьому його внесок можна порівняти з внеском нейтронного
опромінення.
1. ASTM Standard E 1921-97. Standard Test M ethod fo r Determination o f
Reference Temperature To fo r Ferritic Steels in the Transition Range,
Annual Book o f ASTM Standards, Vol. 03.01 (1998).
2. All-Union State Standard 25.506-85. Strength Analysis and Tests. Methods
fo r M echanical Testing o f Metals. Determination o f Crack Resistance
(Fracture Toughness) Characteristics in Static Loading [in Russian],
Gosstandart USSR, M oscow (1985).
3. Report on the Irradiation Conditions, NPP/PCP-4-IRLA(99)-D1 [in Russian],
U nit 5, Novovoronezh (1999).
4. All-Union State Standard 1497-73 (ST SEV 471-77). Metals. M ethods o f
Tensile Testing [in Russian], Izd. Standartov, M oscow (1983).
5. All-Union State Standard 9454-78 (ST SEV 472-77, ST SEV 473-77). Metals.
A M ethod o f Testing fo r Impact Bending at Low, Room, and Elevated
Tempeatures [in Russian], Izd. Standartov, M oscow (1982).
6 . Standards fo r Strength Analysis o f PNAE G-7-002-86 Nuclear Pow er Plant
Equipment and Pipelines [in Russian], Energoatomizdat, M oscow (1989).
7. Steinstra D. I. A., Stochastic M icromechanical M odeling o f Cleavage
Fracture in the Ductile-Brittle Transition Region , M M 6013-90-11, Ph.D.
Thesis, Texas A& M University, College Station (1990).
8 . W allin K., “A simple theoretical Charpy V — K Ic correlation for irradiation
em brittlem ent,” in: A SM E Pressure V essels and P iping C onference,
Innovative Approaches to Irradiation Damage and Fracture Analysis, PVP,
Vol. 170, ASM E, New York (1989).
9. W allin K., “The scatter in K Ic results,” Eng. Fract. Mech., 19, No. 6 ,
1085-1093 (1984).
10. Ballesteros A., Strizhalo V. A., Grinik E. U., et al., “Determ ination of
reference tem perature T0 for steel JRQ in an unirradiated state and
construction o f a M aster Curve,” Probl. Prochn., No. 1, 41-51 (2002).
Received 05. 04. 2002
ISSN 0556-171X. Проблеми прочности, 2003, № 1 47
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