Radiation-Induced Embrittlement of WWER-440 Reactor Pressure Vessel Steel under Loading

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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’c h u k ,a V. A. S triz h a lo ,b L. S. N o v o g ru d sk ii,b A. B a lle ste ro s,c L . D e b a rb e r is ,d F. S ev inid

a Institute for Nuclear Researches, National Academy of Sciences o f Ukraine, Kiev, Ukraine

b Pisarenko Institute o f Problems o f Strength, National Academy of Sciences o f 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 ь ч у к а, В. А. С т р и ж а л о 6, Л . С. Н о в о г р у д с к и й 6, А . Б а л л е с т е р о с в, Л . Д е6a р6ер и сг, Ф . С е в и н 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. К лю чевы е слова: нейтронное облучение, радиационное охрупчивание, кор­ пусная сталь, “M aster curve”, механическая нагрузка, давление теплоносителя. In tro d u c tio n . A t the tim e being, the radiation em brittlem ent o f the reactor vessel m aterials is evaluated b y the results o f the C harpy-type surveillance specim ens or fatigue pre-cracked com pact tension specim ens. T he surveillance specim ens are irradiated in the herm etically closed container assem blies w ithout stress. The reactor vessel m aterial is actually subjected to the pressure o f the coolant.



F or determ ining the effect o f m echanical load under irradiation on the em brittlem ent rate o f the reactor vessel m aterials tw o container assem blies w ere irradiated in com m ercial unit. T hey w ere com pleted w ith the specim ens cut o f four various grades o f steel, having been utilized for m anufacturing reactor pressure vessels in the form er Soviet U nion. Three kinds o f specim ens w ere prepared o f each type o f steel, nam ely:

- static tensile specim ens;

- C harpy-V type im pact specim ens;

- T-L-oriented stressed and unstressed com pact tension specim ens.

The specim ens o f each steel w ere located in three layers. There w ere 2 tensile, 2 C harpy-V and 4 CT (2 stressed and 2 unstressed) specim ens in every layer o f each assembly. Thus, each assem bly w as com pleted w ith six stressed and six unstressed 1/2T com pact tension specim ens m anufactured o f each type o f vessel steel.

Irradiation w as perform ed on such a level o f the reacto r core th at w ithin one year o f irradiation to accum ulate the dose com parable w ith that accum ulated by the reactor core in the center o f the active zone during the design service life (40 years).

The prim ary m aterial property, determ ining its susceptibility to spalling, is the ability o f the m aterial to resist the propagation o f a crack. F or reactor steels the ductile-brittle transition tem perature shift in the irradiated state has to be defined as com pared to the unirradiated one, therefore the application o f large- size surveillance specim ens is unacceptable.

E m ploying the M aster Curve approach for ferritic steels enables to obtain credible values o f the critical stress-intensity factor Kjc in the w ide tem perature range testing sm all-size specim ens [1]. A ccording to the recent standardizing docum ents, M aster C urve is constructed on the basis o f the results o f testing sm all-size specim ens at a single tem perature such that steady grow th o f a crack before the beginning o f the fracture is practically precluded. This allow s to sim plify considerably the procedure o f defining j -integral and use experim ental basic approach o f nonlinear fracture m echanics.

The p aper presents the results o f determ ining the reference tem perature TO and constructing a M aster C urve proceeding from the experim ents, carried out for the steel 15Cr2M oVA (base m etal o f W W ER -440 reactor vessel m etal) in irradiated, unirradiated, and irradiated stressed states.

M a te r ia l a n d S pecim ens. C hem ical com position o f the investigated steel is presented in Table 1. 1/2T CT specim ens w ere studied. Fatigue cracks (a 0) on them w ere initiated in line w ith the requirem ents o f the State Standard 25.506-85


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 m etal o f an operating W W E R -type reacto r is u nd er the pressure o f the coolant. In W W ER -1000 it equals 16 M Pa, that results in the stress o f the


Radiation-Induced Embrittlement o f WWER-440 Reactor

vessel m etal o f about 173 M Pa. F or im itating such stresses m echanical load was created on CT specim ens by m eans o f the loading screw and m etallic bellows. Stressed specim ens w ere assem bled into tw o chains, linked w ith tw o bellows, next to unstressed specim ens, enabling the correct com parison o f the data for both groups o f specim ens.

The design o f the assem blies allow ed w ater o f the prim ary circuit to w ash the specim ens, so their irradiation tem perature w as equal to the tem perature o f the coolant, i.e., ~ 2 9 0 oC.

Fast neutron fluences (E > 0.5 M eV ) accum ulated b y the studied specim ens in experim ental assem blies are calculated b y the K urchatov Institute in the fram ew ork o f the P roject TACIS PC P -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


7 9.15 •Ю19 9.24 •1019

8 1.04 -1020 1.05 •Ю20

9 1.16-1020 1.17 •Ю20

M e th o d s o f Testing. Static tensile testing w as conducted in accordance w ith the State Standard 1497-73 [4]. R em ote-controlled tensile-testing m achine installed in the “h o t” cham ber w as applied as experim ental equipm ent. The error o f the breaking stress determ ining w as ± 2 5 M Pa. The active grip m oved w ith the speed o f 1 m m /m in. Tests w ere perform ed at room tem perature and at 350OC. K eeping tem perature w as w ith the accuracy ±2OC. The m axim um error o f determ ining strength characteristics is 5% and th at o f plasticity - 2%. The results o f static tensile tests are show n 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 , a m P R M A>,% A S %% 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 -5 9 -31 -46

Note. Over the line are given results for unirradiated specimens and under the line - for irradiated specimens.

Fracture toughness tests w ere carried out in line w ith A ST M Standard 1921-97 [1] on the testing electrom echanically-driven m achine Instron-8500 w ith the critical load 100 kN. The m achine is rem ote-controlled and it is installed in


the “h o t” cham ber o f the Institute for N u clear R esearch o f the N ational A cadem y o f Sciences o f U kraine in the fram ew ork o f the program TACIS PCP-IV. The procedure o f m easuring tem perature, load and crack opening satisfied the requirem ents [1] that provided determ ination o f the above characteristics w ith high accuracy and obtaining reliable P — V diagram s.

Since crack opening w as m easured on the front surface o f a specim en, according to p. 7.1 [1] the correction for displacem ent o f the loaded line was applied by m ultiplying the m easured values b y 0.73.

Fracture toughness testing tem perature w as found as

w here C = —28°C for 1/2T specim ens [1]. The tem perature T28j w as defined by the results o f C harpy im pact testing ( 1 0 X 1 0 X 55 m m ) on the pendulum ham m er K M D -30D w ith im pact accum ulated energy 300 J in the tem perature range (—80°C )-(+ 100°C) in line w ith the standard [5] (Fig. 1). The processing o f im pact to u g h n ess d ata (E ab) w as co n d u c te d ac co rd in g to [6] b y m ean s o f the approxim ation function o f hyperbolic tangent, the equation o f w hich looks like

w here E ab is the absorbed energy (J), A is the average value o f E ab betw een the m axim um (E max) and m inim um (E min) values, B = (E max — E min ) / 2 , Tk0 is the tem perature corresponding to the value A, and C is the em piric constant. Param eters A, B , C , and Tk0 are defined b y the processing o f the experim ental points b y the m ethods o f least squares.

D uctile-brittle transition tem perature for unirradiated specim ens (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.


Radiation-Induced Embrittlement o f WWER-440 Reactor

From the data o f Fig. 1 the fracture toughness test tem perature =

— 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 tem perature T = — 90°C and

T = — 40°C for unirradiated and irradiated specim ens, respectively.

O b ta in e d R esu lts. As a result o f testing seven unirradiated specim ens at — 90°C P — V diagram s, characteristic for cleavage cracking, have been got. By the results o f processing these diagram s the values o f /-in te g ra l w ere defined as the sum o f elastic and plastic com ponents. The values o f the stress intensity factor (K J c ) w ere calculated from the j c values obtained for each individual specim en. The values o f J c and K j c for unirradiated specim ens o f steel 15Cr2M oVA are illustrated in Table 4.

P — V diagram and the fracture surface in first irradiated specim en at the

testing tem perature Ttest = — 40°C testify the considerable stable grow th o f a crack. The value o f K j c for this specim en is equal to 129.95 M PaV m (Table 4). A ccording to the recom m endations o f C SR IEM “P rom etei” on the assessm ent o f fracture toughness in W W ER -440, W W ER -1000 reactor vessel m aterials, the testing tem perature w as low ered b y 20°C, but at Ttest = — 60°C the stable crack grow th w as also m ore intensive. A t Ttest = — 80°C cleavage cracking took place in the irradiated specim ens. T herefore all the rest irradiated (stressed and unstressed) specim ens w ere tested at Ttest = — 80°C. The results o f testing all the specim ens are show n in Table 4. N one o f the K j c values should be qualified as all the values do no t exceed K j c(limit) value for the above testing tem perature.

M a s te r C u rv e A p p ro a c h A p p lic a tio n . 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 specim ens [7, 8]. The position o f the curve K j c(T ) on the tem perature coordinate is established from experim en tal d eterm in atio n o f the tem p eratu re T0 at w h ich th e m ed ian v alu e K j c( med) for 1T specim ens is

100 M P a V m . Ferritic steels are k now n to be very heterogeneous due to the orientation o f individual grains as w ell as to heterogeneities o f the grain boundaries. C arbides and different non-m etallic inclusions on the grain boundaries m ay be nucleus o f m icrocracks. T heir random location relative to the crack front in the m aterial conditions the large spread in values o f fracture toughness testing sm all-size specim ens.

A s it is show n b y W eibull [9] the curve shape for ferritic steels w ith yield strength from 275 to 825 M P a is determ ined by the index o f the exponential curve

b = 4 and K min = 20 M PaV m . These data are received on the basis o f the statistic

analysis o f a huge num ber o f experim ental data, obtained b y various researchers all over the w orld on the specim ens o f different sizes.

A ccording to the requirem ents [1], w hose adequacy has been confirm ed in [10], the values obtained on specim ens o f size 0.5T w ere recalculated into the values equivalent to those obtained on specim ens o f size 1T by the form ula

= T

35J/cm2 for unirradiated specim ens w as — 62°C and for irradiated ones it was



T a b l e 4

Results of Fracture Toughness Testing of Base Metal 15Cr2MoVA

No. State 3 P KJc (1/2T), MPaVm KJc (1TH MPaVm K 0, MPaVm K , KJc (med) • MPaVm 0 ^ E-? 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 w e calculated the scale param eter

K 0 = N ( K Jc(1T)) i=1 N - 0.3068 1/4 + 2 0, (4)


Radiation-Induced Embrittlement o f WWER-440 Reactor 15 C r2M o V A u n irra d iated IV I .a 2 3 4 KJc (i) - Km

M aster 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)


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) = (K0 - 2 0 )[ln 2]1/4 + 20. (5)

The m agnitude o f the reference tem perature (70) w as determ ined using the follow ing expression:

t - T - 1 i ( Kjc(med)

0 test 0.019 n ,


70 (6)


The values o f the param eters calculated according to form ulas (3 )-(5 ) for three states o f steel 15K h2M FA are listed in Table 4. M aster Curves (Fig. 2) w ere constructed by the equation [1]:

Kjc(med) = 3 0 + 7 0 e x p [0 .0 1 9 (r- r 0 )]. (7)

F or visual representation o f the results o f testing, W eibu ll’s m odel w as used according to w hich the probability o f fracture is evaluated by the form ula

P f = 1 - e x p { -[(K jc - K min ) / ( K 0 - K min ) f }, (8)

w here K min = 20 M PaV m and b = 4.

Figure 2 dem onstrates that the data o f K Jc lie w ithin ± 5 % confidence interval o f M aster C urve approach.

M ean-square deviations T0 estim ated according to [1] are equal ± 7 °C for unirradiated specim ens and ±6°C for irradiated ones in both states.

Figure 3 illustrates M aster C urves for three states o f steel 15Cr2MoVA: unirradiated (1), after irradiation unloaded (2), and irradiated loaded (3). R eference tem perature T0 for irradiated unstressed specim ens exceeds its value for unirradiated specim ens. Total effect o f neutron irradiation and stress results in the increase o f the reference tem perature in this case b y 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 onclu sio n s. The effect o f neutron irradiation on radiation em brittlem ent o f N i-free vessel steel 15Cr2M oVA is studied in tw o states: und er m echanical load and w ithout 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 tem perature o f the fatigue pre-cracked specim ens o f steel is discovered. The above effect is revealed in the reference tem perature T0 for stressed irradiated specim ens being high er than for irradiated unstressed ones. The effect o f the stressed state caused b y m echanical load,


Radiation-Induced Embrittlement o f WWER-440 Reactor

im itating the pressure o f the coolant, on the ductile-brittle transition tem perature is qualitatively com parable w ith the effect o f neutron irradiation. From the physical point o f view, this phenom enon is conditioned b y the h ig her rate o f the form ation o f radiation defects due to the total effects o f neutron irradiation and stress. Р е з ю м е Н аведено результати визначення еталонної тем ператури Г0 та побудовано “M aster curve” на основі експериментів, проведених для сталі марки 15Х2МФА (основний м етал корпусу реактора ти пу В В Е Р-440) в неопром іненом у, опро­ м іненом у й опром іненом у під навантаж енням стані. П оказано, що механічне навантаж ення, яке ім ітує ти ск теплоносія, прискорю є радіаційне окрих- чення, при цьом у його внесок м ож на порівняти з внеском нейтронного опромінення.

1. A STM Standard E 1921-97. Standard Test M eth od f o r Determ ination o f

R eference Temperature To f o r F erritic Steels in the Transition Range,

A nnual B ook o f A S T M Standards, Vol. 03.01 (1998).

2. All-Union State Standard 25.506-85. Strength A nalysis and Tests. M ethods

f o r M echanical Testing o f M etals. D eterm ination o f Crack R esistance (F racture Toughness) C h aracteristics in S tatic L oadin g [in R ussian],

G osstandart U SSR , M oscow (1985).

3. R eport on the Irradiation C onditions, N PP/PC P-4-IR LA (99)-D 1 [in Russian], U nit 5, N ovovoronezh (1999).

4. All-Union State Standard 1497-73 (ST SEV 471-77). M etals. M ethods o f

Tensile Testing [in R ussian], Izd. Standartov, M oscow (1983).

5. All-Union State Standard 9454-78 (ST SEV 472-77, S T SEV 473-77). M etals.

A M eth od o f Testing f o r Im pact Bending a t Low, Room, and E levated Tempeatures [in R ussian], Izd. Standartov, M oscow (1982).

6. Standards f o r Strength Analysis o f PNAE G -7-002-86 N uclear P o w er Plant

E quipm ent and Pipelines [in R ussian], E nergoatom izdat, M oscow (1989).

7. S teinstra D. I. A ., Stochastic M icrom echanical M odelin g o f C leavage

F racture in the D uctile-B rittle Transition R egion, M M 6013-90-11, Ph.D.

Thesis, Texas A & M U niversity, C ollege Station (1990).

8. W allin K., “A sim ple theoretical C harpy V — K Ic correlation for irradiation e m b rittle m e n t,” in: A S M E P ressu re V e sse ls an d P ip in g C o nference,

Innovative A pproaches to Irradiation D am age and Fracture A nalysis, PVP,

Vol. 170, A SM E, N ew Y ork (1989).

9. W allin K., “The scatter in K Ic resu lts,” Eng. Fract. M ech., 19, N o. 6, 1085-1093 (1984).

10. B allesteros A ., Strizhalo V. A ., G rinik E. U., et al., “D eterm ination o f reference tem perature T0 fo r steel JR Q in an u n irrad iated state and construction o f a M aster C urve,” Probl. Prochn., N o. 1, 41-51 (2002).

Received 05. 04. 2002




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