Investigation of samples taken from Kozloduy unit 2 reactor pressure vessel

NIId~l[ ELSEVIER

amll~sign Nuclear Engineering and Design 160 (1996) 59-76

Investigation of samples taken from Kozloduy unit 2 reactor pressure vessel A. Kryukov a, P. Platonov", Ya. Shtrombakh", V. Nikolaev u, E. Klausnitzer', C. Leitz~, C.-Y. Rieg a aRussian Research Center, Kurehatov Ins:itute, Moscow, Russia t'Central Research b:stitute, Prometey, St. Petersburg, Russia :Siemens AG, Power Generation Group (KWU), Erlangen, Germany aElectricitd de France, Villeurbanne, France

Received 18 July 1994: revised 21 April 1995

Abstract Within the framework of the 6 month W A N O program, small samples were cut from the inside surface o f the Kozloduy NPP unit 2 reactor pressure vessel to assess the actual condition of the pressure vessel material before and after annealing. The actual values of the weld metal characteristics required for estimating radiation-limited lifetime--the ductile-to-brittle transition temperature (DBTI') in the initial state (Tko) and the phosphorus and copper contents which affect the radiation stability of steel--wer,: not dete.,~qined during manufacturing. The Kozioduy unit 2 pressure vessel had no sm'ceillance program. Radiation stability was evaluated using dependencies based on analysis results for surveillance samples taken from other VVER-440 reactors. For this reason, the actual pressure vessel characteristics and their changes in the course of reactor opere*ion, as well as comparison of experimental with calculated data were the principle objectives of the study. Instrumented impact tests were carried out on sub-size specimens of base and weld metal. Correlation dependencies were used with standard tests to determine DBTTs for the base and weld metal (in accordance with Russian standards): base metal before annealing 40 °C, after annealing 16 °C; weld metal before annealing 212 °C, after annealing 70 °C. The estimated value of Tko, for the initial, unirradiated weld metal, was 50 °C. The experimental results were compared with a prediction of the extent of radiation-induced embrittlement of Kozloduy unit 2 pressure vessel materials. It was confirmed that radiation-induced embrittlement of the base metal does not impose any limits on the radiation-limited lifetime of the pressure vessel. The predicted increase in the DBTT of the weld metal as a result of irradiation (about 165 0(2) is practically equal to the experimental result (162 °C). However, the value of Tf obtained from tests before annealing (212 °C) is about 40 °C higher than the estimated value, i.e. the calculation does not produce a conservative estimate. This was explained by a low estimate of Tko (10 °C), which had been calculated using data from chemical analysis of the weld metal, performed by the manufacturer. The investigations on the samples, however, yielded an estimated value of Tko = 50 oC. The effectiveness of annealing in restoring the mechanical properties of irradiated W E R - 4 4 0 reactor pressure vesse!s was confirmed. Recovery annealing lowered the DBTT of the weld metal by 85% or more of its radiation-induced shift.

0029-5493/96/$15.00 © 1996 Elsevier Science S.A. All rights reserved S S D I 0029-5493(95)01066- I

60

A. Krvu,~ov el aL / Nuclear Engineering and Design [60 ([996) 59-76

1. Introduction Within the framework of the 6 month WANO program small samples were cut from the inside surfaee of the Kozloduy NPP unit 2 reactor pressure vessel in February-March 1992 to assess the actual condition of the pressure vessel material. This investigation was undertaken by a consortium of Eiectricit~ de France (EdF) and Sien-,ens AG with MOHT Otzhig acting as subcontractor. The detailed study was carried out in hot cell laboratories of the RRC Kurchatov Institute and the CRI Prometei. The report presents the results of investigations carried out at the Kurchatov Institute and Prometey with specialists from Siemens AG and EdF providing methodological support and participating in the analysis of experimental results. Unit 2 of Kozloduy NPP (VVER440 type, Fig. 1) was commissioned in 1975 and, by the time that the samples were cut, it had been in operation for 16 campaigns, including three campaigns with partial core loading (with shielding assemblies). The coolant temperature at the inner surface of the pressure vessel is 270 °C. Unfortunately, the actual values of the weld metal characteristics required for estimating radiationlimited lifetime--the ductile-to-brit.tle transition temperature (DBTT) in the initial state (Tko) and the phosphorus and copper contents which affect the radiation stability of steel--were not determined during manufacture. Therefore, calculated values of these characteristics obtained from empirical dependencies had to be used for the radiation-limited lifetime estimates for the weld metal. The Kozloduy unit 2 pressure ves~l had no surveillance irradiation program. Radiation stability was evaluated using dependencies based on analysis results from surveillance samples taken from other WER--4zt0 reactors (Amayev, 1993a). For this reason, the actual pressure vessel characteristics and their changes in the course of reactor operation, as well as comparison of experimental with calculated data were the principle objectives of the present study. There was one more important task, namely to provide experimental proof of the effectiveness of thermal annealing of the pressure vessel, performed in March 1992, to restore the mechanical properties of the steel.

Therefore, samples were cut before and after annealing. 2. Sample description Fig. 2 shows schematically how the samples were cut from the reactor pressure vessel. A total of 15 samples were cut out, 12 of which (four base metal and eight weld metal samples) fulfilled the size requirements and were accepted. Seven sam-

427O

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Fig. I. Schemeof w:i0 ~eamlocationsin VVER-440reactor pressure vessel.

A. Kryukov et al. / Nuclear Engineering and Design 160 (1996) 5 9 - 7 6

61

1,2,3 weld ne'bxt before eJmealir'~ 6,7,0 bose r,e~t before aaneaUn9 # 4,5 weld neCal o¢'ter" ameatln9

H H VESEL ~

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Fig. 2. The scheme of cutting samples from Kozloduy unit 2 reactor pressure vessel. pies were taken before annealing and five after. The experimental plan is shown in Table 1. The samples were to be used for impact testing to determine the value of DBTT of both materials before and after annealing. A special thermal treatment was used to simulate a state ,of the weld metal similar to that of the initial (unirradiated) state. Tensile tests were planned to provide strength and plastic characteristics of the weld metal before and after annealing. Fabrication of sub-size specimens for testing was performed in several stages. Workpieces and Table I Experimental plan for Kozloduy unit 2 samples No. of samples

Base metal

Weld metal

Total Before annealing After annealing For impact tests

4 2 2

8 5 3 6

4

then plates were manufactured from the samples, and then specimens were produced from the plates. Fig. 3 shows the scheme of sample cutting to fabricate both types of specimen (5 × 5 and 3 x 4 mm 2) for impact testing.

3. Spectrometric analysis of weld and base metal composition The main components and impurities were determined by spectrometric analysis tbr every sample tested. The analysis was carried out using a Baira Speetrovac FSQ optical emission spectrometer. Primary and secondary standards for low alloy steels were obtained from the Bureau of Analysed Samples Ltd. (UK). In the case o f weld metal, the weld and crown were examined separately. The specimens for spectroscopic analysis were plate-shaped with a thickness of about 5 m m

A. Kryukov et al. / Nuclear Engineering and Design 160 (1996) 59- 76

h) o f the pressure vessel; in addition weld metal specimens were tested after annealing at 560 °C f o r 2 h. The base metal hardness was: -irradiated condition, 21.5 _+2.5 HRC; - - annealed 475 °C, 150 h, 17.0 _ 1.0 HRC. The following hardness was obtained for the weld metal: - - irradiated, 20.5 _+ 1.0 HRC; - - annealed 475 °C, 150 h, 13.5 + 1.5 HRC; - - annealed 560 °C, 2 h, 15.04- 1.5 HRC.

5. Impact testing The tests were performed using an instrumented RKP-300 impact machine with a potential energy o f 80 J and a pendulum impact velocity o f 3.87 m s - ~. The pendulum impact machine has a Charpy tup, compliant to DIN, with tensometric sensors. The signal is analyzed to yield the force-displacement or force-time function. Data read from Table 2 Results of phosphorusand copper analysis in weld and base metal Fig. 3. Schemeof samplecutting to fabricate sub-sizeimpact specimens(3 x 4 and 5 x 5 mm2). for the weld samples and about 4 mm for the base metal samples. The contents o f P and Cu for each sample are listed in Table 2.

4. Hardness measurements Hardness measurements were performed with a diamond cone at a load o f Pr = 150 kgf using a standard TK-2M hardness meter. Hardness was measured in at least three locations on the surface opposite the notch in impact specimens. Measurements were made on two specimens from each sample. To obtain valid data the measurements were performed so that the indentations were spaced at least 3 mm apart. Hardness was measured on base and weld metal specimens before and after the recovery annealing (475 °C for 150

Material

Element content (wt.%) P

Cu

5a-b-BW 5a-b-WC 4-a-BW 4-a-WC 3b-a-BW 3b-a-WC 7-b-BW 7-b-WC 1l-b-nW ! I-b-WC

0.0354 0.0363 0.0345 0.0373 0.0347 0.0342 0.0418 0.0374 0.0371 0.0377 0.0340 0.0354 0.0318 0.0373 0.0260-0.0376 0.0350 0.0178 0.0162 0.0166 0.0177

0.178 0.i 66 0.! 7i 0.I 83 0.175 0.167 0.155 O.!72 O.167 0.175 0.176 0.169 0.162 0.183 0. i 62 0.157 0.174 0.163 0.163 0.167

12-aoBW

12-a-WC 6-b-BW 6-b-WC 10-b-BW

10-b-WC lb 2b 8a 9a

BW, weld base; WC, weld crown.

A. Kryukov e! al. / Nuclear X~ngineering and Design 160 (1996) .99-76 "'[

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TEHPERFITURE F i g . 4. A b s o r b e d

energy as a function

of temperature,

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base metal before annealing,

9 x 10 m n cm--"

(E > 0.5

MeV). both tile instrumented tup and from the angle of rise of the digital angle indicator were used to determine absorbed energy. The absorbed energy was read off the angle indicator. The temperature dependencies of absorbed energy and lateral expansion for base metal and weld metal are presented in Figs. 4-13.

6. Determination of weld metal mechanical properties by tensile testing

MPa at 270 °C. The effect of annealing on the reduction of ultimate tension strength is less pronounced (approximately by 100-120 MPa). No visible effect of annealing on plastic properties was observed.

7. Validation of correlation dependencies between DBTT deterndned for sub-size and Charpy specimens Z 1. D B T T

Tensile tests were performed on two specimens from each of sample nos. 6 and 12 at 20 and 270 °C. The results are tabulated in Table 3. The data indicate that the weld metal investigated had rather high strength properties before annealing (Re2 = 765 MPa, R m = 830 M P a at room temperature) and rather high plastic properties. Annealing caused a reduction in yield strength, on average by 120 MPa at room temperature and 140

criteria for standard

specimens

In accordance with A S M E code (ASME, 1992) the crucial values o f absorbed energy are 41 and 68 J, as well as a lateral expansion of 0.9 m m for Charpy specimens. According to Russian guidelines (Energoatomizdat, 1989) the crucial levels for absorbed energy are 48 and 70 J. From these the critical temperature Tk is determined as follows: if TTo -- T4s ~< 30 °C, then T4s is assumed as Tk,

A. Kryukov et al. / Nuclear Engineering and Design 160 (1096) 59-76

64

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TEHPERATURE

3 x 4 m m 2 s p e c i m e n s , b a s e m e t a l b e f o r e a n n e a l i n g , 9 x 10 ' s n c m - 2 ( E > 0 . 5

MeV). otherwise if 7"7o- T4s/> 30 °C, t h e n T~o -- 30 °C is assureed as Tk. In addition, the F~ussian guidelines stipulate that at the temperature o f Tk + 30 oC the lateral expansion should n o t be less t h a n the critical value (0.9 m m for a C h a r p y specimen). Otherwise, a temperature 30 °C lower t h a n that at which t~e critical lateral expansion is achieved, is assumed aS

Tk.

7.2. Definition of DBTT criteria for sub-size specimens T o establish the relevant correlations between sub-size and standard specimen test results, the experimental d a t a for 3 x 4, 5 x 5 and 10 x 10 nun 2 cross-section specimens of grade 1 5 K h 2 M F A steel (six heats) a n d welds (seven welds) with different impurity contents were summarized and analyzed. Specimens o f two materials

supplied by the I A E A ( J R Q a n d JPI) were also tested. Part o f the material was irradiated and annealed. Both s t a n d a r d a n d sub-sized specimens were irradiated under identical conditions. T h e irradiation temperature was 270 °C, the fast neut r o n fluence for the specimens was 1 x 1020 n c m - 2 ( E > 0 . 5 MeV). T h e transition temperature in the impact tests is determined from a fixed level o f absorbed energy. It is assumed that similarity o f the criterion m a y be insured only w h e n this level is constant in relation to the energy o f fully ductile fracture (upper shelf); that is, w h e n A. = constant USE

(I)

where A~ is the critical level o f the a b s o r b e d energy (J) and U S E is the upper shelf o f the a b s o r b e d energy (J). Statistical treatment o f the d a t a base d e m o n strated the following relation between the upper shelf levels f ~r 5 x .5 and 10 x 10 ram 2 specimens:

65

A. Kryukov et aL / Nuclear Engineering and Design 160 (1996) 59-76

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a n d for 3 x 4 a n d 10 x 10 m m 2 specimens: USE(10) = 22.3 USE(3)

(3)

where the n u m b e r in parentheses is the characteristic dimension o f the specimen cross-section. U n der this consideration, tl~e critical absorbed energies corresponding to 41, 48, 68 a n d 70 J for s t a n d a r d specimens m a y be assumed equal to 5, 6, 8.5 a n d 9 J (in r o u n d figures) for 5 x 5 m m 2 specimens. In accordance with Eq. (3) for specimens of 3 x 4 m m 2 cross-section, the critical absorbed energy corresponding to 48, a n d 70 J should be 2.2 a n d 3.3 J (in r o u n d figures). According to A h l s t r a n d et al. (1993) the critical absorbed energies for 3 × 4 m m 2 cross-section specimens are 1.9 a n d 3.1 J. These levels corres p o n d to 41 a n d 68 J for standard specimens.

200

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specimens, base metal after annealing, 0 × 10.9 n¢m - 2

(E>

0.5

T h e lateral expansion o f the specimen is a n o t h e r critical value for the determination o f T~. Specific experimetit~ were performed to support this criterion for 5 × 5 a n d 3 × 4 m m 2 cross-section specimens. T h e first step taken in the course o f this work was to determine the lateral expansion corresponding to 50% shear f r a c t m e for s t a n d a r d specimens. This was a n old Russian criterion for transition temperature determination. T h e lateral expansion a n d fracture appearance o f irradiated a n d unirradiated specimens o f 1 5 K h 2 M F A steel a n d welds were analyzed; the analysis covered a total o f 432 d a t a points. A 50% shear fracture corresponds to a lateral expansion o f 0.93 m m for s t a n d a r d C h a r p y specimens, which is practically identical to the criterion o f 0.9 m m in Western standards ( A S M E , 1992). Following this, the dependence o f lateral expansion o f s t a n d a r d specimens on absorbed energy was determined (Fig. 14). A total o f 571 experimental points were measured. There are n o groups o f d a t a points which

66

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Fig. 7. Lateral expansion as a function of temperature, 3 x 4 MeV); sample no. 8.

ram 2

c o u l d be regarded as outliers in t h e irradiated o r u n i r r a d i a t e d conditions, neither for t h e b a s e m e t a l n o r for t h e weld metal. Fig. 15 p r e s e n t s t h e dependence o f lateral e x p a n s i o n o n a b s o r b e d energy for 5 x 5 m m 2 cross-section specimens. U s i n g the accepted p r o c e d u r e for d e t e r m i n i n g t h e critical lateral e x p a n s i o n , a v a l u e o f 0.35 m m wae d e t e r m i n e d for 5 x 5 m m 2 cross-section specimens. T h e d e p e n d e n c e o f lateral e x p a n s i o n for 3 x 4 m m 2 cross-section s p e c i m e n s w a s d e t e r m i n e d in a n a n a l o g o u s m a n n e r a n d w a s f o u n d to be 0.3 m m (Fig. 16). T h e v a l u e o b t a i n e d is identical to t h e lateral e x p a n s i o n values established for W e s t e r n pressure vessel materials ( A h l s t r a n d , 1993). Z3. Dependence o f transition temperature on specimen dimensions T h e relationship between t h e D B T T o f 10 x 10 a n d 5 x 5 m m 2 s p e c i m e n s in t h e irradiated a n d u n i r r a d i a t e d conditions, d e t e r m i n e d f r o m a statis-

specimens, base metal after annealing, 9 x 1019n era-2 (E > 0.5

tical t r e a t m e n t o f t h e experimental d a t a ( m e n tioned in Section 7.2) c a n be p r e s e n t e d as follows (Fig. 17): D B T T ( 1 0 ) = D B T T ( 5 ) + 46 °C

(4)

where D B T T ( 1 0 ) is t h e transition t e m p e r a t u r e determined from Charpy specimens and DBTT(5) t h e transition t e m p e r a t u r e d e t e r m i n e d f r o m 5 x 4 x 27.5 m m 3 specimens. T h e m e a n s q u a r e deviation 21 °C. It w a s suggested t h a t t h e final t e r m be r o u n d e d u p to 50: D B T T ( 1 0 ) = D B T T ( 5 ) + 50 °C

(5)

T h e correlation between D B T F f r o m C h a r p y a n d 3 x 4 x 27 m m 3 s p e c i m e n s h a s been studied in A h l s t r a n d et al. (1993). A n u m b e r o f low alloy steels with different h e a t t r e a t m e n t were used, s o m e o f w h i c h h a d been irradiated, a n d also weld materials. T h e results o f t h e statistical analysis s h o w e d a linear relation between t h e c o m p a r e d D B T T values with a slope o f u n i t y a n d a n intercept o f 65 °C:

A. Krvukov et al. / Nuclear Engineering and Design 160 (19967 59-76 3

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TEMPERATURE

Fig. 8. Absorbed energy as a function of temperature, 5 × 5 mm2 specimens, weld metal before annealing, 6.7 x 10'9 n crn -z (E > 0.5 MeV).

DBTT(10) = DBTT(3) + 65 °C + 30 cC

(6)

For unirradiated and irradiated 10 × 10 and 3 x 4 mm 2 specimens of steel heats, mentioned in Section 7.2, practically the same relationship was found (Fig. 18): DBTI'(10) = DBTI'(3) + 65 °C

(7)

with a mean square deviatioa of 24 °C.

8.

D~ussion

Tables 4 and 5 list the critical D B T r s for *he materials under investigation as stipulated by P.uss:.an and Western standards. According to the Russian standard, the critical transition temperature of weld metal determined from sub-size specimens is - - before annealing Tk = 162 °C -after annealing Tk = 20 °C

--

after heat treatment (560 °C, 2 h) Tk = --24 °C Concerning the estimated value of T~o (the value o f Tk in the unirradiated condition), it was shown (Amayev, in press; Leitz, 1993) that after brief annealing at temperatures of 550-560 °C the Tk of the annealed metal is by 2 0 - 3 0 °C lower than Tko. In this case, the value for weld metal would be 0 °C. It was repeatedly demonstrated (Amayev, 1993b) that following brief thermal treatment of irradiated steel at 460-490 °C, Tk is not more than 2 0 - 3 0 °C higher than Tko- Thus, for weld metal, Tko is in the range of - - 2 4 to + 20 °C, and it is advisable to assume Tko = 0 °C for subsequent calculations. Table 6 presents the values o f Tk for standard Charpy specimens obtained using the correlation between the results of impact tests of standard and sub-sized specimens. It was d ~ m e d necessary to compare the experimental r~:-.aits with the calculated values for the weld metal and also to

68

A. Kryukov et aL / Nuclear Engineering and Design 160 (1996) 59-76

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............!...............! .................i .................t .................!..............i............i ........ i........................................................... | ; ~ i i i i i i i t ........... ,, ................. ) ................. ) ................. ! ................. ! .............. ~ ..... t ................................... , .................. ~............ J i i i i

F ~

9

[

.............

~

~

i

..................' .................i............-

i ........................ /~ i~.................. ~i.................. ,i............ i ~i................. ~!................ ~~ ................. i ~ ................. ~i................. ~).............. ~. | , $ /,f,. × i~ "~ i i i ~ i : I ..................................... 1................................................... T ....... o i ............. ~ ............. ~ . o. Samplel~Sa . . -) ~ ..o"/~ -

" ..........................................

,,,i

-:~00

i .................. L ~

.... ~ , ~ , , i

- 100

O

................ ~............ ': ....

~............ ~.. x s ~ o

........

) ....

i ....

I~0

20~

I ....

, ....

N, ~

..--

, .... :

300 CEL TEHRERRTURE

Fig. 9. Lateral expansion as a function of temperature, 5 × 5 mm~ specimens, weld metal before annealing, 6.7 × l0 ~ n mn -~ (E > 0.~ MeV).

predict the dcgrcc o f radiation induced embrittlem e n t for operation o f K o z l o d u y unit 2 following annealing. A s far as the ba~e metal is concerned, radiation embrittlement does n o t place any constraints o n pressure vessel lifetime. As stipulated by the Russian standards, the critical temperature o f irradiated VVER-A40 reactor pressure vessel materials is determined in the following manner: Tf = Tko + ATf ,= Tko + .4iF I/3

(8)

where P and C u are the p h o s p h o r u s a n d copper contents (%) o f the metal• As was noted above (see Section 1), the values o f Tko a n d the p h o s p h o r u s a n d copper contents for weld metal were n o t determined during m a n u facture o f the K o z l o d u y unit 2 reactor pressure vessel. Calculations were used to estimate these values. Specifically, a formula for determining Tko from the contents o f some alloying a n d impurity elements present in the weld metal was proposed:

where Tf is T~ in the irradiated condition, F is the fast n e u t r o n fiuence with E > 0.5 M e V (n c m - 2 ) and Af is the radiation embrittlement coefficient, d e p e n d e n t o n the copper and p h o s p h o r u s content o f steel. According to Russian guidelines, the estimate o f Af for weld metal is obtained from the equation:

F o r the Kozloduy unit 2 reactor weld metal the formula yields:

Af ----800(P + 0•07Cu)

Tko=

(9)

rko = ]01.6-- 171(% MnX% M o ) + 15! .8(%Mn)(%V) + 8224(%Si)(%P) -- 42139(%S)(%P) - 2726(%P) -

163(%Si)(% M o )

lO°C

(10)

A. gryukov et al. / Nuclear Engineering and Design 160 (1996) 59-76 -""

Jl ....

~.o 113

,,

i

I ....

I ....

i

!

I ....

I ....

I ....

i

:

:.

...................................................

r ~''

'1 ....

6

............. i................. i .... [..........* ............. i ........... i....... ; ...... .

4

-.

•

~%,,,i,,,,i~--

-100

i

....

O

i ....

l~gl

i ....

.~.....

i

~ ....

i

69

i

..i ..i i ;

.............................................................

,,,i

t .......

i i.......

12

o -2~0

I ....

~ .... 2~0

i

i............. i............

osm,o3 .............

31~

CEL

TEHPERRTURE

Fig. i0. Absorbed energy as a function o f temperature, 5 x 5 mm 2 specimens, weld metal after annealing, 6.6 x 1019 n cm - 2 ( E > 6.5

MeV). Phosphorus content was determined based on phosphorus content of the weld wire, while for copper the maximum copper content known at the time was assumed. This resulted in values of P = 0.036%o and Cu = 0.21% for the calculation. This yields Af = 40.6. The calculated fluence incurred by 1991 was 6.7 x 1019 n cm -2. The calculated value of Tr by 1991 for the Kozloduy unit 2 pressure vessel weld was 165 °C. Therefore, the calculated estimate of the weld DBTT before annealing was Tr= 175 °C. This value is substantially smaller than the experimental result of Tf-- 212 °C. It was necessary to determine the reasons for this higher experimental value compared with the calculated one. As for the degree of radiation-induced embrittlement, the experimental value of A T f = 162 °C (Table 6:212 - 50 -- 162 °C) is practically equal to that calculated (165 °C). This is owing to the fact that the calculated and the experimental

phosphorus contents are sufficiently close (calculation: P = 0.036%, Cu = 0.21%; experiment: P = 0.037%, Cu = 0.17%). The experimental value of neutron fluence (6.68 × 1019 n m - 2 ) is practically identical to that calculated. The comparison of experimental and calculated Tko values calls for the following comments. Owing to the presence of a large number of free variables not very precisely determined in Eq. (10), the reliability of the Tko values obtained from Eq. (10) is not very high. A higher experimental value of Tto, compared with that calculated, was also obtained from the analysis of samples cut out of the Novovoronezh unit 3 reactor pressure vessel (Amayev, in press). The calculation used Tko = 5 °C, and analysis of the samples yielded Tko = 55 °C. For the Kozloduy unit 2 weld metal the calculated value of Tko = 10 °C is 40 °C lower than the experimental one (50 °C). This in turn gives a low calculated estimate of the weld Tr (175 °(2) corn-

A. Kryukov el al. I Nuclear Engineering and Design 160 (1996) 5 9 - 7 6

MM

"~ . . . . . .

1.3

-

1.2 1.1 I

•

I .... I ........

i

I........

~ ..........

)

.11 ,?

i~

-

I

!

:

:

- ........ i ............. ~................ ~................. )............... I................. ~................. !.............................................. i............................ -. : i ~ ! i i i i i ] I - ...... i................................ l i ................................ l......................................................... ~.............................. : I [ ! ~ i ! I : . ~ ! I i i i i i i :

: ........... i ............. i............... i............. i................. !............. !

.9

I ....

~

!~ ..................... I i.......i............................................... [ ! ~i............ tm............ ~r

- ........... i ......................

: •

i

i.......

i

! .... i ......... i............... !................. )............ .... i............. i ............................. •................ I............ -"

................................................................................ i ~ i Y ~ ! i ~ ! J" i

i |

I

,

: '

- ........... i ................................................................. i..................... :7 ~ .........i............... i................. t................ )............ -

:

.B

~

I

!

i

/

!

!

.

,4

- ............ !....................................................................... h Z i ........................ i................. i............ : ! I i i /ii i i i ! : -.."............. i................. !............... I................ T;~-~ ...... : .......i................ i.................................. i.......... : ........... i i i ! I ~ : i ~ i i i ~ ....................... ~................ t............................. iJ...... ~ ........, ................ ~............................................................

.3

-: ........ 'i..........i................. ]~............... :./.i...... i............ I............i

.5

:

,2 .1

!

,

i

7,

i

i

~

:

i! ........................... i ! i}.................. i1........... --"~

: ........... !............... i..........i............. .......:~-.i: ................ !........... i........ i.............. io~.,,.,-~

i x~,....~

:-.......... i .............................. ~ .................................................... -,,,,I

-aeo

....

,~,-,-r~,, 1 ~ , ,

-100

t ....

B

i ....

i ....

100

i,,,

M..~.,,

200

i ....

I,,,,:1

300 CEL TENPERRTURE

Fig. I!. Lateral expansion as a function af temperature, 5 × 5 m m (E > 0.5 MeV).

2

pared with the experimental value (212 °C). It can be seen from the data in Table 6 that the weld metal Tk recovered by 142 °C, which corresponds to about 85% of its radiation-induced shift. This result agrees with the predicted value. For the change in weld transition temperature during reirradiation following annealing, the Russian guidelines give the following equation:

is 1.62 × 10 ~8 n cm - 2 ( E > 0 . 5 Me'*/). Therefore, the shift in T k for the Kozloduy unit 2 reactor pressure vessel weld is calculated from the formula:

Tf---- Tko + 20 ° C + A f F I/3

(l !)

where the value of Ar is assumed to be the same as during the first irradiation and F is the fluence incurred during reirradiation. In order to determine Af, it is necessary to use the results of chemical analysis of sample metal (the maximum values are P = 0 . 0 3 7 5 % and Cu =0.18°/o) to assess the change in Tr following annealing. In this ease, Eq. (8) yields A t = 40. When shield assemblies are installed, the annual increment o f fluence picked up by the weld metal

specimens, weld metal ~fter annealing, 6.6 x lO~9 n

Tr = 70 + 40(1.62n) u3

em-2

(12)

where n is the reactor vessel service, life in years. Fig. 19 shows the transition temperature as a function o f neutron fluence before annealing (prediction compared with measurement) and after annealing (predicted with Eq. (12)). However, the estimate yielded by Eq. (12) is a rather conservative one. Presently, much experimental data indicate that the rate of radiation-induced embrittlement during reirradiation is lower than during the initial irradiation. However, in order to change the standards, it is necessary to collect additional data, including those on radiation-induced emkfittlement of pressure vessel metal operated for several years after annealing,

A. Kryukov et al. / Nuclear Engineering and Design 160 (1996) 59-76

3"

"'1 .... I ................

[........

..............

24

.............i . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

?i~

................................................... , ........... *. . . . . . . . . . .

il

~, . . . . .

i

i

I .... I .... I .......

ii

' :~. . . . . . . . i.................~..................~.............

i !

' ~

i............ i...............................

10

......

i j~,.~......................................... : ...... .............................................. ,,,i

-200

....

~ , T

-100

....

I ....

0

, ....

i ....

100

!:......

~ ....

~...........................................

i ....

i ....

i ....

200

,

....

300

' ....

::

CEL

TEMPERATURE

Fig. 12. Absorbed energy as a function of temperature, 5 × 5 mm2 specimens, weld metal after annealing 560 *C, 211;6,6 x 1019n cm -2 (E>0,5 MeV); sample no. 11. 9. Conclusions

Investigation of the samples cut out of the Kozloduy cnit 2 reactor pressure vessel established the following. (1) Instrumented impact tests were carried out on sub-size specimens of base and weld metal. Correlations were used to determine DBTTs for base and weld metal corresponding to standard specimen tests (in accordance with Russian standards): base metal before annealing 40 °C, after annealing 16 °C; weld metal before annealing 212 °C, after annealing 70 °C. The estimated value of Tk, corresponding to the initial, unirradiated state, was 50 °C.

(2) Tensile tests of weld metal showed sufficiently high plastic properties in the state betbre and after annealing. The reduction in yield strength as a result of annealing was 120 MPa. (3) The experimental results were compared with a prediction of the extent of radiation-induced

embrittlemem of Kozloduy unit 2 pressure vessel mate-!als. It was confirmed that radiation-induced embrittlement of the base metal does not impose any limitations on the radiation-influenced lifetime of the pressure vessel. The predicted increase in DBTT of the weld metal as a result of irradiation (about 165 °C) is practically equal to the experimental result (162 °C). However, the value of Tc before annealing (212 °C) obtained from tests is about 40 °C higher than the esthnated value, i.e. the calculation does not produce a conservative estimate. This was explained by an understated value of Tko (10 °C), which had been calculated based on a chemical analysis of the weld metal performed by the manufacturer. The results of investigations on the samples, however, gave an estimated value of Tko = 50 °C. (4) The effectiveness of annealing as a means of restoring the mechanical properties of irradiated W E R - 4 4 0 reactor pressure vessels was confirmed. The DBTT of the weld metal was reduced by no less than 85°,6 of its radiation-induced shift.

A. Kryukov et al./ Nuclear Engineering and Design 160 (1996) 59-76

72

HH 1.3

.....

1.2

-- ............ i................. i................ i................. i ............... i................................ i.............................. . ~ { i i i i i i i ............. i................. i................. ~................. i................. i............... ! .................................

1.1

I ........

I ..............

~Tr"

- ............ J.................. ].................. J.................. i ................. ~...........

i i ............. ~................. i................. ~................ ~................. i........

I .9

............. i.................. i.................................. '................. i..........: i

.0

i

i

~

'i .... i ...........

!

!

i

i

....... i

: ,,

i

i .............i......

'i ................... Ii............. :

................................................ i............. i.................. i............... 2i.....i................................. i................. i............. z

'?

............. i ................................... i................................. i'~"i

~.s

.............

_1

.............. i ......................... i ................................ i .............

............. 1.................. !.................. i.................. Lo,.,,~ ........... i .......... i ............... i i I~.,,T_,,,,.~' i i oq -

.2

~i................. ~i................. i ................. "............... ~.............. ~

~

............. i................ |...................................... ~...........

i

i

~

•- .' .i . ............. . . II................

i

....i....~

.... '

:

~ ........ ~

i

i

...... I................................... i............................. i.

i I i ~ ............................................................ i................. i............ i........~

0 -2OO

i

i.......... i

i

i

i .....i .......... ~ ...........

i

i

i

i

.........~ ............. i............

....i.... i........ i....i....i....i....i,,,i 0

-100

I00

200

300 CEL TEHPERRTURE

Fig. 13. Lateral expansion as a function of temperature, 5 × 5 m m 2 specimens, weld metal after annealing, 5f0 °C, 2h; 6.6 × 1019 n c m - 2 (E > 0.5 MeV); sample no. II.

Table 3 Mechanical properties of weld metal Specimen no.

Test temperature (°C)

R,~ (MPa)

Rpo.2 (Mea)

St, (MPa)

.4 (%)

Ap (%)

Z (%)

6-I

23

823

753

1395

20.8

14.4

58.8

6-2

23

834

779

1480

21.9

14.3

60.8

12-I

23

736

627

1465

20.4

12.2

64.3

12-2

23

711

638

1255

18.3

I 1.8

60.5

6-3 (>.4 12-3 12-4

270 270 270 270

747 706 620 595

671 647 530 507

1255 i ! 20 1075 980

12.4 15.2 14.0 18. I

10,2 I 0.0 8.8 8.8

53.3 53.6 57.2 55.2

Sk is the true fracture stress.

.4. Kryukov el al. / Nuclear Engineering and Design 160 0996) 59-76

73

Z.6

1 o oo,.r b... : : , . , 0-irr

base meut~

..o ~

0

z~,~rtrs~

. . . . . t.~.'. . . . . . . . . . . . .

o

O

O

/w

~t~"~"

-

P:'rr.;:',~..T~..

so

1oo 15o 200 25o ABSORBED ENERGY, J Fig. 14. Correlation between lateral expansion and absorbed energy (standard specimens).

O - unirr O- in. ,, - a r t + a n n Ik . . . . p l e a of

/

/

~0.5

n ~ -

Yc

I~

o.o O

6

I0

/USE(IO)

U~'~" 16

|0

2fl

30

3S

40

ABSORBED ENERGY. J Fig. 15. Correlation between lateral expansion and absorbed energy (5 x 5 mm 2 specimens).

A. Kry,kov et aL / Nuclear Engineering and Design 160 (1996) 59- 76 |.0'

~

o-unirr bue metal o-irr base metal t,-lrr+ann base metal +-baae metal of H o v o - V o r o n e z h RPV

o.e

~0.6

l Utj

0

÷

~ ~

~ 1 ~ + / " ~_ 0.0

[~l$[~(lO) _A 7O ~

z

o-irr weld metal ~x -- il rr r + aannnn w ee ll dd m e t a l

4 8 ABSORBED E~ERGY° J

10

8

Fig. 16. Correlation between lateral expansion and absorbed energy (3 x 4 ram 2 specimens).

JO D

1.6o unirr o ~rr ro leo

/

i

"g

~,~/~

i f,,,,

-1o0:

r

-~ze,

...........

- \ Dur(tOl:De'rr(s)÷.c

"-

-'~

.....

-;,5 ........

5x5

DBTr,

C

~ .........

deg

!;) . . . . . . . .

i

Fig. 17. Correlation between full-size Charpy and sub-size (5 x 5 mm 2) speci,men transition temperature.

Table 4 DBTT for 3 x 4 x 27.0 mm base metal specimens Material condition

Sample no.

Critical temperature (°C) Russian standard

Before annealing After annealing

I, 2 8

2.2 J

3.3 .I

-25 - 49

- 12 - 34

Western standard 0.3 mm -44 - 65

1.9 J

3.1 J

-29 - 53

- 14 - 37

75

,4. Kryukov et al. / Nuclear Engineering and Design 160 (1996) 5 9 - 7 6

/

M ~

A o

,~1

unirr in"

o

/ /

60

o

O

O:

~.~.

0~

r _~o ) . ~ / , ,

-lO0-i~ti . . . . . . .

~ D~jTT~"l O}=DB'i'F(3} +65c

n

"--~'6. . . . . . .

"~ ........

30 . . . . . . . . .

~0. . . . . . . .

i

3x4 DBTT, deg C Fig. 18. Correlatic,n between full-size Charpy and sub-size (3 × 4 mm2) specimen transition temperature.

Table 5 DBTT for 5 x 5 x 27.0 mm specimens of weld metal Material condition

Sample no.

Critical temperature (°C) Russian standard

Before annealing After annealing After 560 °C, 2 h

5, 7 3, 4

II

Western standard

6J

9J

0.35 mm

5J

8.5 J

91 11 - 24

163 50 - 1

192 44 - 12

73 -2 - 32

150 40 - 5

Table 6 Estimates of DB'I-r of Kozlodui unit 2 pressure vessel (PV) materials Condition

DBTT (°C) Base metal

Before PV annealing After PV annealing Before start of plant

Weld metal

Russian standard

Western standard

Russian standard

Western standard

48/70 J

41 J

68 J

48/70 J

41 J

68 J

40 16

36 12

51 38

212 70

123 48

200 90

-

-

50

-

-

76

A, Kryukov et al. / Nuclear Engineering and Design 160 (1996) 59- 76

25O

200

..-~'"'""" o°°'°

150

i

~

.s"

.°"

~,

[

[ /

/.. " "

-

/

r-

/' .......... n

:/

"

i;

........

t00

.........

50~

0 O,OOn+O

1992

3,00o+19

6,00e+19

year

Tfpredicted(TKO=IO) Tfpredicted(TKO.50) Tf measured annealingeffect "If afterannealing (conservative~ediclion)

2007

9,00n+19

NmJtmnRue~elcm~-2l(E:.O.SMeV) Fig. 19, Tr as a function of neutron fluence (and time).

Acknowledgments The w o r k described it: this article was conducted by order o f the Committee o f Energy o f Bulgaria and was financially supported by the C o m m i s s i o n o f the E u r o p e a n C o m m u n i t y in the frame o f the P H A R E program. The helpful cooperation o f Kozloduy N P P is also acknowledged.

References

R. Ahlestrand, E. Klausnitzer, D. Lange, C. Leitz, D. Pastor and M. Valo, Evaluation of the Recovery Annealing of the Reactor Pressure Vessel of NPP Nord (Greifswald) Units I and 2 by Means of Subsize Impact Specimens, ASTM STP 1170, ASTM, Philadelphia, 1993, pp. 321-343. A. Amayev, A. Kryukov, V. Levit and M. Sokolov, Radiation

Stability of WWER-440 Vessel Materials, ASTM STP 1170, ASTM, Philadelphia, 1993a, pp. 9-29. Amayev, A. Kt'yukov and M. Sokolov, Recovery of the Transition Temperature of Irradiated WWER-440 Vessel Metal by Annealing, ASTM STP 1170, ASTM, Philadelphia, 1993b, pp. 369-379. A. Amayev, V. Badanin, A. Kryukov, V. Nikolayes, M. Rogov and M. Sokolov, Use of Sub-size Specimel,s ."or Determination of Radiation Embrittlement of Operating Reactors Pressure Vessels, ASTM STP 1204. ASTM, Philadelphia, in press. ASME Boiler and Pressure Vessel Code, Section ?,, Appendix G, Section II, Appendix E, 1992. Energoatomizdat, Standards for strength calculations of components and piping of nuclear power plants, Energoatomizdat, Moscow, 1989. C. Leitz, E. Klausnitzer and G. Hofmann, Annealing experiments on irradiated NiMoCr weld metal, Effect of Radiation on Materials: 16th Int. Symp., ASTM STP 1175, ASTM, Philadelphia, 1993, pp. 352-362.