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Irradiation experiments in the testing nuclear power plant VAK
Nuclear Engineering and Design 72 (1982) 65-79 North-Holland Publishing Company
65
IRRADIATION E X P E R I M E N T S IN THE TESTING NUCLEAR POWER PLANT VAK J. F O H L
Staatliche Materialpriifungsanstalt (MPA), University of Stuttgart, Stuttgart, IV. Germany Ch. L E I T Z
Kraftwerk Union AG (KWU), Erlangen, W. Germany and D. A N D E R S
Forschungszentrum Geesthacht (GKSS), IV. Germam' Received 19 March 1982 Within the German Research Programme "Integrity of Components" the first two capsules were irradiated in the Testing Nuclear Power Reactor VAK. The materials are of the 22 NiMoCr 3 7 and 20 MnMoNi 5 5 types and represent the lower bound of the base material regarding upper shelf energy and chemical composition (Cu, S, P), as well as a state of material which does not meet both chemical and toughness requirements (low upper shelf test melt). Tensile, Charpy, drop-weight, and fracture mechanics specimens were irradiated up to a range of 1.5 to 2 × 10t9 cm 2 ( E >1 MeV). Despite the materials being at or beyond the specification limits, the results show irradiation sensitivity which can be predicted from the US Reg. Guide Trend Curves (1.99) and KWU Trend Curves in a conservative manner. The procedure to determine the adjusted reference temperature RTND-r (adj.) on the basis of AT41J (following ASTM E 185) could also be confirmed as conservative by comparing the different criteria derived from Charpy and drop weight tests in the unirradiated and irradiated condition. The results of fracture mechanics testing in the linear elastic range show a remarkable temperature margin to the Kl:curve of ASME XI. Prestrained compact tension specimens CT 40 mm made of 22 NiMoCr 3 7 material with an upper shelf energy of approx. 100 J were wedge loaded in a range up to 30 MPa m 1 / 2 and exposed to the water environment during radiation. Macroscopic examination gave no indications of stress corrosion cracking. From tests of these specimens in the linear elastic range, a fracture toughness K~'~, which was not affected by the prestrain and environment history, was found depending only on the overload applied during the prestraining procedure.
1. Objective of investigation N e u t r o n radiation causes a change in the mechanical properites of materials. The influence expected during the complete time of operation according to the conservative trend curves is taken into account for the design calculation of reactor pressure vessels. With the help of surveillance specimens which are irradiated in a position of higher flux than at the inner surface of the pressure vessel wall, the actual change in properties is experimentally determined so that the conservativism of the assumption can be confirmed. For new plants built according to the basis safety concept [1], only a little change in toughness is to be expected. This is based on the present trend curves due to the restricted chemical composition and the limita002%5493/82/0000-0000/$02.75
tion of fluence by 1 × 1019 cm -2. Therefore, the surveillance programmes of such plants include only a minim u m number of Charpy and tensile specimens to validate sufficient material toughness for the whole time of operation. Vessels of which the material has higher sensitivity against neutron embrittlement due to chemical composition and which are exposed to a higher neutron fluence cannot be valuated alone on the basis of Charpy energy in case of a low toughness level. Each single case requires a complete fracture mechanics analysis. The radiation of the necessary large specimens in the respective power reactor is usually not possible. The research p r o g r a m m e "Integrity of C o m p o n e n t s " (F KS) is focussed on the following tasks - the lower toughness level at which safe operation
© 1982 N o r t h - H o l l a n d
66
J. F6hl et al. / Irradiation experiments in the testing nuclear power plant VAK
can still be assured, how to transfer results from small scale specimen testing to the c o m p o n e n t behaviour, - which criteria describe the material toughness and whether the behaviour of low toughness material achieved by metallurgical and thermo-mechanical processes represent the behaviour of neutron embrittled material. The aim of the irradiation experiments of the F KS is to determine fracture mechanics properties of irradiated materials and to establish the limit of possible irradiation embrittlement even for extreme material states in the as delivered conditions. In order to obtain results in a short period of time, most of the irradiation experiments are carried out at high neutron flux in research reactors, since large specimens can be irradiated only there. Such short-time irradiation experiments involve additional parameters, such as neutron flux and neutron energy spectrum. Therefore, further corrections are necessary for applying the results to the pressure vessel wall [2]. The differences in neutron spectra for a variety of reactors and surveillance positions are shown in fig. 1, representing the influence of neutron spectra on the primary damage rate (displacement per atom = dpa). The damage rate by neutrons with an energy E > 1 MeV in the capsule of the research reactor F R G - 2 (CK 120 type capsule b e h i n d a steel shield) a m o u n t s to approx. 30%, in the V A K radiation position approx. 55%, and at the RPV wall (inner surface) of a 1300 M W power reactor approx. 80%. The main portion of material damage is included, if only the damage caused by neutrons in the energy range E > I MeV of a power reactor is calculated. However, this portion is only
power reactors ~ > lO0
i
~
~
~
""
b
-
% 50
0
neutron energy
~
=
a
b
b
~--
o
b
~_~IMeV
• [] ~ []
E~3MeV 3~-E>1MeV 1~-E~0,1MeV E~-O.1MeV
o =surveiU, pos. b : RPV inner surface
z so o ~
~_
_
0
Fig. 1. Comparison of neutron spectra of research and power reactors.
approx. 1 / 3 in a research reactor. Therefore, the effective change in material properties is greater after irradiation in the research reactor than at a comparable fluence E > 1 MeV in a power reactor. The comparison of the spectra a m o n g power reactors at the pressure vessel wall shows that the differences are not so severe. There are differences of about three orders between n e u t r o n flux at the RPV-wall of a 1300 M W reactor and the irradiation position in the FRG-2. The current available knowledge about the effect of dose rate on the material embrittlement is not yet sufficient. However, it can be assumed that high flux radiation experiments cover the embrittlement of reactor pressure vessels in a conservative way, since the time is shorter for recovery during radiation. The F KS irradiation p r o g r a m m e includes a great n u m b e r of various irradiation positions in different reactors as well as a wide range of materials from which those single parameters can gradually be assessed. The irradiation experiments conducted in the Testing Nuclear Power Reactor Kahl (VAK) are an i m p o r t a n t link in this chain of transferability of results from research reactors to surveillance programmes, and finally, to assess the safety of the reactor pressure vessel. Furthermore, it is possible to investigate the superposition of parameters during radiation as they are cyclic loading, water environment, temperature and neutron radiation. This paper presents the results from the first F KS irradiation experiment carried out in the VAK.
2.
Materials
For the first VAK-irradiation experiment, the following materials were selected: - KS 01 forged shell 22 N i M o C r 37, fig. 2; lower limit of specified upper shelf energy (80-100 J). KS 12 forged shell 20 M n M o N i 55, fig. 3, manufactured after M H K W * procedure at the b o u n d of the specified chemical composition. KS 07 B forged slab, fig. 4, with chemical composition exceeding the specified limits of 22 N i M o C r 37 "Low Shelf Test M e l t " (LSTM) with an upper shelf energy of 3 5 - 4 0 J. T h e Charpy energy curves of the materials investigated are represented in fig. 5 together with material as currently used in commercial plants (KS 13). * In the MHKW process a trepan is removed from the ingot and the core is subsequently refilled by electroslag remelting (ESR).
C
Si
Mn
P
Chemical composition (wt %) S
~
KS07
0.18 2Q0.28
1.27
148
1.15 1.55
0.62
0.71
) belt line region.
032
p-~
Ks12
KS 13
a guiding value, (
0.10 0.35
0.29
0.24
0.15 0.25
20 MnMoNi 55 Optimized
~
KS01
~-~
<~0.012
[ ~
~
0.007 0.007
~
~<0.012
~
~
22 NiMoCr 37 Specification 0.17 0.50 ~<0.025 ~<0.025 after VdTUV 0.25 ~0.35 1.00 365 (4.72) Optimized ~<0.20 0.20 a 0.85 a ~<0.008 ~<0.008
Material
Table 1 Materials of the 1. VAK irradiation experiment
0.30 0.50
Ni
0.50 0.60 0.80 1.00
Mo
0.04
015
~<0.20
0.49
0.41
0.45 0.72
FCZiq063
0.40 0.45 0.55 0.85
11.-~0.74
['~0.95
<~0.40 ~<0.55 1.20 a
Cr
0.046
0.01 0.04
~<0.05
AI
0.05
~
~<0.12 (~<0.10)
002
~<0.02
0.05
0.012
<~0.01
<~0.05
V
0.022 <0.01
0.027
0.010 0.040
10.---~ ~<0.003
[~]
~<0.10
~<0.20
Cu
N
As
0.007
~<0.011
0.012
0.01
~<0.013
-
-
<0.05
<0.01
~<0.005
Sb
-
Co
<0.005
0.007
0.011
~<0.030 ~<0.030
0.16
_
~<0.030 ~<0.030
Ta
0018 <0.005 <00,
~<0.025
0.026
0.017
<~0.010 ~<0.013 ~<0.015
Sn
~
~
~.
5
~"
5"
~"
t,,
z2
68
J. Fehl et al. / Irradiation experiments in the testing nuclear power plant VAK
-,~ 550
6000
_
KS 01
K S 07 B
22 NiMoCr37
22 NiMoCr 31
/
- mid -ptane
.f,t.b
/q/I
heod
0 , ~S
Fig. 2. Forged shell KS 01, upper shelf energy 80-100 J.
--
LJ/
Fig. 4. Forged slab KS 07 B, low shelf test melt (LSTM) with upper shelf energy of 30-40 J.
KS 12 20 MnMoNi55 ~
318o
ESR
J
unirrodieted
,
~,"
so
-200
/ I- .....
:
l~/
l~
t
-I00
,'
't
+IL-TI "°°' ~KsoTB
.._ .........
0
100
2_
KS NDT-T "C 01 +5 07B (*30) f2 con~ +10 12 ESR -40 13 -20
200
_/.~
300 °C 1,00
temperoture
Fig. 5. Range of material toughness selected for the irradiation experiments in the research program "Integrity of Components" (F KS).
Fig. 3. Forged shell KS 12 after M H K W procedure, lower bound of chemical composition.
Table 2 Tensile properties of the unirradiated materials
specification
•p0,2 N/ram2
R
FIT
350 ° C
RT
350 ° C
RT
390
343
560 70O
505
19
16
45
35
438
379
586
552
25
20
55
45
m
N/ram 2
z%
A5% 350 ° C
RT
350 ° C
22 N i M o C r 37 KS 01 K S 07 B
i1 21 5481
[13
8,3]
20 M n M o N i 55 K S 12 c o n v °
535
433
694
604
20
15
56
46
627
23
21
67
62
550
27
20
73
65
KS 12 E S R
558
436
705
K S 13
437
375
577
69
J. FOhl et al. / Irradiation experiments in the testing nuclear power plant VAK
The chemical composition and the tensile properties are compiled in tables 1 and 2, respectively, together with the specified data of steel 22 N i M o C r 37 (ASTM A 508 C1 2) and 20 M n M o N i 55 (ASTM 533 B C1 1, 508 CI 3). All materials were used in the quenched and tempered condition and simulated post weld heat treated. It has to be taken into consideration that in case of KS 12 it was not possible to remove specimens from the usual quarter thickness location, since the ESR zone has only a thickness of 60 to 80 mm. Therefore, the specimens represent the material state of a highly quenched surface region.
Table 3b Specimens of the 1. VAK irradiation experiment, capsule B1
I K S 01
3.1. Specimen and capsule preparation
The investigation including preparation of specimens and capsules as well as testing in the hot cells was carried out by KWU-Erlangen. Type and number of specimens for the first irradiation experiment are listed in tables 3a and b. Wedge loaded compact tension specimens from KS 01 material were exposed to the water environment to find out if simultaneous radiation will cause stress corrosion cracking. In order to separate effects caused by prestrain from those related to the water environment, wedge loaded specimens were also placed inside the capsule. Due to the geometrical conditions at the capsule position it was only possible to apply compact tension specimens with a thickness of 40 mm (CT 40). Three different stress intensity levels, namely, apTable 3a Specimens of the 1. VAK irradiation experiment, capsule AI capsule
KS 12 conv.reglon KS 12 ESR-region KS 07 c o m o a r i son m,~ter i a l 1)
I = low
m
A 1
Drop W e i g h t P2
Charpy ISO-V 1) fluence
I
f l u e n c e 1)
I
h
I
m
h
18
6
6
6
18 18
h 32)
8
K S 12 cony, region
8
K S 12 ESR-region KS 07 comparison
r~q ]
8
9
6
B
8
16
material
m -medium2
-
10
19
cm
-2 h
=
high
not emcapsulated
3. Irradiation experiment
KS 01
OT 40
f l u e n c e 1)
I = low
material
Oharpy
tensile
material
18
3.2. Irradiation o f the specimens
18 18
m = medium 2 " 1019cm -2
prox. 10, 20, and 30 MPa m I/z, were chosen. To apply the stress by inserting the wedge, the specimens were overloaded by 60 to 1()0% because of the compliance of the specimen-wedge system. Prior to installing the specimens in the reactor, they were exposed to the water environment at 280°C in an autoclave for 14 days so that a magnetite layer could be developed to protect the specimens against corrosion in the cold reactor water from the time of installation until the final heat up to operational temperature. Subsequent loading of the specimens in a tensile machine to achieve partial unloading did not indicate relaxation during the exposure in the autoclave. Two thin walled capsules ("collapsed cans") were used so that good heat transfer from the specimens to the coolant was assured. The arrangement of the specimens in the capsules (AI and B1) is shown in figs. 6 and 7, [2]. For the assembly plan of the capsule A1, the axial flux gradient was taken into account by gradually displacing the specimens towards the core. The corrosion specimens were in a cage through which the water had access to the prestrained specimens. Monitor plates were used in three axial planes to determine the neutron fluence and the temperature during radiation. Each monitor plate contained nine sets of fluence monitors (Fe and Nb), and five sets of melting cones covering the temperature range from 271 to 302°C, fig. 8. Moreover, additional single monitors were placed at the top and the bottom of the capsule.
h = higlq
The Testing Nuclear Power Plant in Kahl is of a boiling water reactor type with a power capacity of 16 MW. At the core barrel, there are three positions at
70
J. F6hl eta/.
/
Irradiation experiments in the testing nuclear power plant V A K
radiation direction
radiation direction
Fig. 6. F KS irradiation experiment in the Testing Nuclear Power Plant Kahl (VAK), capsule AI.
which capsules of the size up to 100 X 150 x 950 m m can be irradiated, fig. 9. A relatively high y-radiation leading to a heat up of the specimens is effective at this location. The reactor was operated at 12 MW, because it is k n o w n that reducing the power to 12 MW, a n d thus, lowering the operation and water temperature will set a n upper temperature limit within the specimen package of 290°C. The actual n e u t r o n flux at this position a m o u n t s to 2.2X 10 ]2 cm - 2 s ] in the centre of the capsule. The n e u t r o n energy distribution can be seen in fig. 1. Approximately 15% of the neutrons have an energy above 1 MeV, however, they cause 55% of the primary damage. In comparison with the spectrum of F R G - 2 , where only 30% of the primary damage is caused by neutrons with an energy above 1 MeV, the V A K spectrum is harder, but not so h a r d as the spectrum of a 1300 M W reactor at the vessel wall. The water conditions i m p o r t a n t to the specimens not encapsulated are compiled in table 4. Both capsules were radiated in the V A K at the same
.
_
Fig. 7. F KS irradiation experiment in the Testing Nuclear Power Plant Kahl (VAK), capsule BI.
time during a period of 100 full power days to achieve the desired fluence of 2 X 1019 c m 2 in the capsule mid-plane.
VAK irradiatioh experiment monitor plate
p0e I"~ ~
temperature monitors.
""~" :
271
zs~ 286
~, i @
÷@@
8
i ~9
288 289
295 ~o2
~
i -
# ;
L
L :
~e e • .@
1~6,5
Fig, 8. Monitor plate with temperature and fluence monitors.
J. F6hl et al. / Irradiation experiments in the testing nuclear power plant VAK
71
10zo cm-2 6 3r
tu 2
-6-
o~ 10~g 8 = ti
capsule ~
1i~*
c
/
capsule Bt
2 101°
Fig. 9, Irradiation positions in the Testing Nuclear Power Plant Kahl (VAK).
0
50 mm distance from capsule inner surface
100
Fig. 10. Comparison of fluence monitors evaluation with calculated data. Table 4 Composition of the VAK cooling water
Oxygen
0,9
50
Chloride
3,5 ppm ppb
pH - v a l u e
5,5
-
6
el. conductivity
17
-
71
HS/cm
4. Evaluation of the irradiation experiment
absolute fluence in axial direction of tile middle axis was derived from all iron detectorsl According to the theoretical gradient in tangential a n d circumferential direction, all fluence data of one plane were transformed to the mid-point of the m o n i t o r plate. F o r each mid-point a mean was calculated using a least square fit. All mid-point fluences were fitted by a smooth curve in axial direction, fig. 11. This was the basis from which the fluence of each specimen could be calculated. The reference point of the fluence of each specimen was chosen as the mid-point of the tensile, C h a r p y and drop-weight specimens a n d as the m e a n crack length ( a 2 4 - a 3 + a 4 ) / 3 at B / 2 position of the compact tension specimens.
4.1. Fluence 4.2. Irradiation temperature The n e u t r o n field at the core mid-plane was d e t e r m i n e d o n the basis of the actual b u r n up state of fuel elements. This was c o n d u c t e d with the two-dimensional transport code DOT-3 using the n e u t r o n crosssection data of E U R O L I B 4 with 53 energy groups. The calculated fluence a n d the results of the m o n i t o r evaluation of iron (54 Fe (n, p) 54 Mn), a n d N i o b i u m (93 N b (n, n') 93 m N b ) are c o m p a r e d in fig. 10. The calculated values are conservative since the radial gradient of the power distribution of the edge fuel elements was not taken into account. The agreement between the fluence data derived from N b a n d Fe is good (15%) near to the core side of the capsule, however, at the rear of the capsule, the evaluation of N b leads to a 50% lower fluence. The
F r o m the melting cones of the six m o n i t o r plates an irradiation tempe~'ature was evaluated by visual examination. In all cases, the 288°C m o n i t o r was not melted but the 271 °C one. A differentiated e x a m i n a t i o n of the m o n i t o r s with melting points of 284°C and 286°C led to an irradiation temperature in the range of 282°C to 288°C. 4. 3. Testing o f irradiated specimens 4. 3,1. Tensile tests The specimens were tested in the unirradiated condition in the t e m p e r a t u r e range from - I00 up to 350°C, a n d in the irradiated condition from - 150 up to 275°C.
72
J. F6hl et al. / Irradiation experiments in the testing nuclear power plant VAK
500~-! mm~- .... I,O0 - -~
ii
~0o
i E
i
200
"5
"° 100
L
° 0,
]
"-
I
I
i
°E O.5.1019crn "2 t5 ~c~ 100f|ue nce!(Fe)~
~ 2
2oo~ x o
I 4 I I L
; I
i t
2L5/
!
I
Itt
I '
III
300
I I I Ill
~00
v
500
7
Fig. 11. Fluence distribution in the capsules in axial, tangential and radial direction.
The yield strength, as a function of temperature was established to evaluate the validity of fracture toughness. General information concerning the change in tensile properties due to irradiation can be achieved by c o m p a r i n g results from testing at room temperature a n d testing at 275°C. It has to be m e n t i o n e d that irradiated specimens were not tested at room temperature but at 0°C, fig. 12. However, this does not affect the comparison, since the change in properties in this temperature range is neglegible a n d c a n n o t even be represented in
~ unirracliatecl
~" 1200 r .£~ Nlmm21
1 RT J
800 1~
the figure. U l t i m a t e and yield strength are slightly higher in the irradiated condition than in the unirradiated one. Reduction of area a n d elongation, which are usually lower after irradiation show, do not show in every case this tendency. The change in these properties is small a n d the differences can p r o b a b l y be traced back to material a n d test scatter. 4.3.2. Charpy impact tests The test temperatures were chosen in a range up to
~ irradinted VAK 1,5-1,8.10 ~9crn2 ~)tested ot 0 °C
275 °C
O:E1201 ~N/mm
RT 11
275 °C
-
"~, KS 01 30
I
KS12 conv. ESR
KS 01
RT 1)
100
,E
.9 E 0
KS 12 cony, ESE
RT 11
%
E KS 01
KS12 cony. ESR
o KS 01
Fig. 12. Change in tensile properties due to neutron irradiation.
KS 12 cony. ESR
275 =C
1
[
J. Fohl et aL
/
Irradiation experiments in the testing nuclear power plant V.4K
275°C in the irradiated condition so that a complete energy temperature curve could be developed including low shelf, transition region and upper shelf. One specimen was usually tested at each temperature step but in the range of the assumed energy of 68 J three specimens were tested at one temperature. The Charpy energy curves for the unirradiated and irradiated specimens of one fluence level are plotted in figs. 13 through 16. The figure shows also the temperature printed at which the energy of 68 J and 41 J, respectively, and the lateral expansion of 0,9 mm, usually the characterizing criteria, is reached. All criteria before and after irradiation are compiled in table 5. The shift in the transition temperature based on 41 J is compared with the trend curves of US Reg. Guide 1.99, fig. 17. All experimental data fall far below the trend curve corresponding to the phosphorous and copper content. The trend curve of KS 07 B material is identical with the upper limit curve. The trend curves assessed by K W U indicate in some regions a lower transition shift than the Reg. Guide curves. They still represent a conservative shift compared to the experimental results, fig. 18. 4.3.3. Drop weight tests O n l y 6 s p e c i m e n s were a v a i l a b l e to d e t e r m i n e t h e d r o p - w e i g h t N D T - t e m p e r a t u r e . T h e r e f o r e , it w a s n o t p o s s i b l e to e v a l u a t e in e v e r y c a s e a p r o p e r l y d e f i n e d transition temperature. The highest temperature (within a r a n g e of 15 K as t h e g r e a t e s t difference), w h i c h c o u l d b e r e f e r r e d to t w o " n o b r e a k " s p e c i m e n s , w a s t a k e n to e s t a b l i s h t h e d r o p w e i g h t N D T . T h e r e s u l t s are also listed in t a b l e 5. T o r e n d e r p o s s i b l e t h e c o m p a r i s o n o f N D T a n d C h a r p y e n e r g y shift at t h e s a m e fluence, t h e
criteria [*C] unirradinted irradiated 200
J
T6ej /,8 106
1"~1j 25
~
i
KS 01, (T_L) i ~ 22 HiMoCr 37
]
68
criteria [*C] unirrodiated irradiated
20 lOB
F q ~
200
J
°~16o
62
T~9,.. 26 111
(L-T
KS0~
-
i
~100 --
-unirr0dioted
&lJ -200
•
~ -I00
a
o
o
! 1,1jO19crn"2 IE~MeVl /
o 0
i 200 temperature I00
J 300 *C /,00
Fig. 14. Charpy-energy of KS 07 B material as a function of temperature in the unirradiated and irradiated condition•
criteria unirradiated
[*el Tsej 30
KS 12 conventional
20MnMoNi 56
t~160
'
.~ O
~ -- ~
-
~
75
• (T-LI
~
Tos~ 10 103
S~ ~T,---- - ~ ] &l J
7
•
J
~
i
unirr0diat
B8J
-200
T~,j 6
113
irradiated
200 j
[,
----JtZ ~/ irrQdlotedVAK| ~ ~ 1~10~scm2{E>]MeV)
-100
0
IO0 200 temperature
300
*C /.00
Fig. 15. Charpy-energy of KS 12 material (conventional zone) as a function of temperature in the unirradiated and irradiated condition.
criteria [*C] T~ej unirradiated -10 irradiated 90
To~mm 25 77
T~. -32 58
To~mm -1~ 81
200 KS12'ESR (T-L) ~ l t = ~ r . ~ J 20 HnMoNi55
I
;~ 100 I
r~ 6 0 , _ - - ~ _ _ _
T~lj /,
T6aj
73
unirrodi(lted
5
~_---/~eirrQdiQtJdVAKi-~t =-vL~ 4
u
irr(ldi~ted VAK[ 2,0.101Scm"z (E>IM I
I
-200 -200
-I00
0
I00 200 temperature
300
*C i,O0
Fig. 13. Charpy-energy of KS 01 material as a funktion of temperature in the unirradiated and irradiated condition.
-100
0
100
temperoture
200
300 *C tOO
Fig. 16. Charpy-energy of KS 12 material (ESR zone) as a function of temperature in the unirradiated and irradiated condition.
unirrad. irrad.
unirrad. irrad.
unirrad. irrad.
KS 12 konv.
KS 12 ESR
KS 07
adjusted fluence. b after [4].
unirrad. irrad. irrad.
KS 01
Material
1,1 X 1019
X
1019 1,6X 1019
2
25 111
81
-- 14
--
20 108
90
-- 10
136
4 62
32 58
100
40
--40
75
121
75
3,0× 1019 2,5 X 1019
113 65
10
45
5
103
5 75
25 68
NDT
1,9X 1019 1,5X 1019
30 110
48 106
T41 j
55
10 93
25 77
T68 j
1,1X 1019 0,9X 1019
2,1X 1019 1,7 X 1019
To.9 mm
Temperature (°C)
Table 5 Comparison of criteria to describe neutron embrittlement
86
95
111
93
83
52
ATo.9 mm
88
100
106
83
80
58
AT68 J
Temperature shift
58
90
95
70
70
43
AT41 j
t,v90 a 80
if70 a \ 65
ff62 a \ 55
if50 ~ 45
F45 ~ \ 4O
ANDT
119
122
136
121
103
67
chemistry relation
AT b
5
e~
t~
2
75
J. F6hl et al. / Irradiation experiments in the testing nuclear power plant VAK
/,00 irrod' temp. 2'08'°C ' after 00'Reg'.Guide 1.99i77 K ---trend curve KSQ1 • VAK-exp.(AT61J} .~. 200 . . . . . . KS 12 cony. * " ~ , a p ~ KS 12 ESR o " ~ ~]~."~ -~S ~ ' /
~u
100 80
~
so
~L_r.~ --
~
t --~-~"
I I
. . . . . . Cu .
• KS 01 K -* KS 12 cony. <1 o KS 12 ESR ~: D KS 078
I
P
.11 .009 17 .015 .26 .022
I I e
E
~
I 0u :0,10 % r p =0,012Ol,
~35
t2ol I
I
o,lo'/.ou]
[
I '/.
3,oo
g
50 -
10
T-L ,, ,, L-T
+ °A~-'~
~ ~o ~ 20 -,~
150
]
. . . . . . . . . .
2x10 ~7
6
8 10~e 2 ftuence
711 t, 6 ( E > I 1,4eV)
t.
I I'l
2
I 3
-
..... 10TM
2
cm -2
0 ~IP=0:012, 10le 2
3
"/*, a,fterKWU t, 5 6 7 8910 TM
|
i J/ t, 5 0 cm"z 10 z°
neutron fluence e (E>IMeV)
Fig. 17. Comparison of transition temperature shift (T4~j) with the trend curves of US Reg. Guide 1.99.
N D T shift was extrapolated on the basis of the slope of the Reg. G u i d e 1.99 trend curves (AT--q~1/2). The results show that the shift in transition temperature measured at a level of 41 J according to A S T M E 185 a n d K A T draft 3203 is almost equal to the drop weight N D T shift of KS 01 and KS 12 ESR material, b u t greater for KS 12 conventional material. The energy criteria for the u n i r r a d i a t e d specimens a n d the shift caused by n e u t r o n irradiation are plotted in fig. 19. T h e adjusted reference temperature RTND T (adj.) of the irradiated material (closed triangle), derived from the reference t e m p e r a t u r e R T N o T* according to A S M E III (open triangle) a n d the t e m p e r a t u r e shift at the 41 J level, is in all cases higher than the drop weight N D T - t e m p e r a t u r e of irradiated specimens. The procedure based on AT41j following A S T M E 185 to determine the reference t e m p e r a t u r e of irradiated materials could be confirmed as conservative. This was already d e m o n s t r a t e d in previous investigations with high copper weld material [3]. The transition t e m p e r a t u r e shift derived from the chemical c o m p o s i t o n (chemistry relation [4]) covers all experimental data d e t e r m i n e d at 41 J in a very conservative way. 4.3.4. Fracture mechanics tests The fracture mechanics specimens were fatigue precracked in the u n i r r a d i a t e d state. The direction of radi-
* The NTNDT was determined in this case from the Charpy energy mean curve. ASME requires the lowest value of three specimens tested at N D T + 3 3 K being at least 68 J for the energy and 0.9 mm for the lateral expansion.
Fig. 18. Comparison of transition temperature shift (T41J) with the KWU-design curves.
ation was so that the u n a v o i d a b l e fluence gradient occurred in the direction of the crack extension, cf. figs. 7 a n d 11. T h e specimens were tested in a temperature range where valid Kit data after A S T M E 399 could be assumed. The fluence of the compact tension specimens a m o u n t e d to approx. 1,5 × 1019 cm -2 ( E > 1 MeV). To verify the validity of fracture toughness, the yield strength was interpolated in the diagram Rp0.2 = f ( T ) . In the u n i r r a d i a t e d condition specimens were tested up to a thickness of I00 m m (KS 01 a n d KS 12 conv.). As the ESR zone of K 12 is limited in thickness direction, it is not possible to use a larger specimen than C T 40 (cf. section 2). Therefore, Kit-testing was re-
material KS 01
fluence
-50
0
temperature 50
100
•
150
2,1.1011¢m-2JE~IHeVIj ~
m.--.*.----.* IlOl KS 12 1,1.1019 conventioneL
I
1,.9.10z9
!
I
3.0.10t9
KS 12 ESR
2,0,10TM
KS O7 B
1,1.10 ly
Fig. 19. Comparison of criteria derived from Charpy impact testing with the NDT-temperature before and after irradiation; the RTNDT was determined from the Charpy-energy/temperatare mean curve.
76
J. FOhl et al. / Irradiation experiments in the testing nuclear power plant VA K
stricted to only a small temperature range. The test results of the unirradiated and irradiated specimens are plotted in figs. 20 through 22. As only a limited number of specimens were available and because of the relatively large scatter, it is not possible to draw a curve through the measure points. The Klc-CUrve and the adjusted Kit-curve (according to A S M E XI) corresponding to R T N o T and RTNt)T (adj.) are given as reference in the plots. The Ki~-range of the conventional zone of KS 12 determined with CT 40 specimens at a fluence of 1.5 × 1019 cm 2 extend almost up to room temperature. The conservatism of the Kt~-curve of A S M E XI could be demonstrated for all three materials in the unirradiated as well as in the irradiated state. The temperature difference between the experimental results and the KI~ curve of A S M E XI amounts to approx. 40 K and more. 4. 3. 5. Testing of prestrained specimens The visual examination of the wedge loaded specimens indicated no crack growth. The specimens were loaded in a tensile machine and the l o a d / d i s p l a c e m e n t curves plotted. The specimens were overloaded up to 70-200% in order to remove the wedges. F r o m the load displacement curves, the remaining stress intensity was determined. It decreased by approx, up to 14% due to relaxation. There was no difference between specimens in the capsule and outside the capsule. After removing the wedges all specimens were tested at - 5 0 ° C and a K ~ or KQ-Value, respectively, was determined, table6. Macroscopic examination of the fracture surface at the crack tip region did not reveal stress corrosion cracking. The Kjc and the KQ values,
KS 01 NiMoCr 37 o unirradinted • irradiated VAK 1,5.101~crnI[E-1 MeV}
22
M Pa
u ' ~ 100
/~
==
~ ZXT~Ij
2 o
=
50~
i
oL___ -200
KIe after ASME Xl / j~unirradiated . adjusted (ATt.ll)
4000 N mm"3n
3000
j
J ladjl 200
---J~adjusted
&O00
Nmm"If;
{AT{,,j)
'
3000
!
2000
RTWO I
RINDT4udjl
i
-200
_
-I00
0
I00 temperature
200
J o
*C
300
Fig. 21. Fracture toughness (linear elastic) of KS 12 material (conventional zone) before and after irradiation in comparison with the Kmccurve of ASME XI.
respectively, as a function of effective prestrain, are shown in fig. 23 (left). Kmc could not be determined with one specimen, since the fatigue crack propagated only on the surface, but not in the centre of the specimen. Ktc and KQ, respectively, show an ascending tendency with increasing prestrain (prestress). More important than the stress intensity during irradiation is probably the maximum load applied during prestraining, which causes blunting of the fatigue crack, fig. 23 (center). The fracture toughness is not influenced up to a Km value of approx. 35 MPa m 1/2 (1100 N m m 3/2). The fracture toughness of the specimens irradiated regularly without prestrain is again shown in fig. 23 (right), (ef. fig. 20),
MPa~rm
:~
100
KS ~2 ESR 20 MnMoNi 55 o unirradioted • irradiated VAK
1,5I0~9cm ~ (E>IMeV) 1o)
c
1000
100 temperoture
Klcafter ASM£ Xl
o
= 5o!: U o RTNot
!
:
j/~ unirradiated
lOOO
I
•
~
/
!
2000
oo o
-100
goO
(*1
i io
I
F
150~
T
150
ISO F - T ! KS 12 conv. 20 MnNoNi 55 M P u ~ ] ounirrudinted • irradiated VAK 100 1'5 101' cm ~ (E'IMeV)
=C
J 0 300
Fig. 20. Fracture toughness (linear elastic) of KS 01 material before and after irradiation in comparison with the Kiccurve of ASME XI.
0i__L -200
-100
•
~ I/T~/~
Kicoffer ASME XI . unirradiated adjusted (Alto))
&O00 N mm
3~;
3000
2000
1000
,T,,o, ,ooj, 0
0
I00 temperature
200
°C
300
Fig. 22. Fracture toughness (linear elastic) of KS 12 material (ESR zone) before and after irradiation in comparison with the Kk.-curve of ASME XI.
77
J. ['Oh/et al. / Irradiation experiments in the testing nuclear power plant V A K
KS 01 22 IfiHoCr 37 test temperature -50 or • in c0psule [-irradioted 1,5.10~g cm-2 • in water Lwith prestrain
150
150
MPo~
~g
~ lOO
irrodioted 1.5 -10~9cm"2
l
i
HP,
e
I00
c
2
5o
i"
.ronge with
l J
overtoed
50 , e ~ I
o
,
25 50 MPo~mm K1 [prestroin)
0
25 50 MPolmm K] (overlood)
-100
"C
100
temperature
Fig. 23. Fracture toughness of specimens irradiated with prestrain (with and without exposure to the cooling water).
together with the results of the prestrained ones. It is evident that, in this case, prestrain and water environment have no effect on fracture toughness.
5. Cyclic loading test The knowledge of crack growth behaviour is necessary for the lay out of reactor components. It is a well established fact that crack growth rate under cyclic loading is higher in water environment than in air. Some specimens were tested occasionally in an autoclave after irradiation. It can be stated from these tests that irradiation has no effect on cyclic crack growth rate [5]. However, it is not known, whether this statement is also true for simultaneous radiation and cyclic loading as there have not been carried out any tests. It cannot be excluded that the coolant (electrolyte) will change during neutron irradiation in such a way that crack growth will be accelerated due to a higher aggressiveness. In addition to the experiments performed already with static load, a test rig is being built at GKSS. This is capable of applying cyclic load to 40 mm thick compact tension specimens, simultaneously with irradiating in the cooling water environment. The cyclic load is generated by an alternating helium pressure in bellows. The load is directly measured at the end of the specimen chain consisting of 5 CT 40 specimens, fig. 24.
% !% % %
(~) C T 4 0 specimens (~) load bellows ~) coupling links ~)
f
supporting frame
(~) force transmitter (~) t r a c t i o n rods (~) Instrument leeds (~) guard plates
• /
~)
spacer block
Fig. 24. Irradiation rig for crack growth investigations.
78
J. FOhl et al. / Irradiation experiments in the testing nuclear power plant VAK
Each specimen is instrumented with a radiation resistent differential transducer to measure the crack opening displacement. The corresponding crack length can be determined from the compliance curves so that it is possible to evaluate complete d a / d N = f ( A K ) curves. The necessary helium pressure as a function of time is generated in a pressure transducer loaded by a mechanical tensile testing machine• A function generator obtaining its signal from the load cell inside the reactor, controls any load-time function at a frequency of approx. 1 cpm and lower (up to 1 cph). A dummy-run with unloaded specimens will serve as a general check of the instrumentation system and its long-time behaviour. Furthermore, the temperature at the crack tip and in the surrounding water will be measured• The first cyclic experiment is to start at the end of 1982.
tt
.o dl
© t.~ i-
6. Summary E ol
.=_
~r',
< > O
.=_
o£
e~
I.
<<<
<
&
<<<
Within the Research Programme "Integrity of Components" (F KS) two capsules were irradiated in the Testing Nuclear Power Plant Kahl (VAK). The materials KS 01 and KS 07 (22 N i M o C r 37) and KS 12 (20 M n M o N i 55) were irradiated and subsequently tested in the hot cells at K W U . The materials represent lower bound conditions concerning chemical composition (KS 12), and upper shelf Charpy energy (KS 01) and both (KS 07 B), respectively. Tensile, Charpy, drop-weight and compact tension specimens were irradiated. In spite of materials that are at or beyond the specification limit, the results show irradiation sensitivity which can be predicted in a conservative manner from the US Reg. Guide Trend Curves (1.99). The procedure of determining the adjusted reference temperature R T N D v (adj) on the basis of 41 J (following A S T M E 185) was also confirmed as conservative• This has been obtained by comparing the different criteria derived from Charpy and dropweight test in the unirradiated and irradiated condition. It was not possible to determine the shift in fracture toughness, because of the limited number of specimens tested in the unirradiated and irradiated condition and scatter in the test results. The results of fracture mechanics testing show a remarkable temperature margin to the Kl~-curve of A S M E XI. Compact tension specimens of 22 N i M o C r 37 with an upper shelf energy of approx. 100 J were wedge loaded in a range of up to 30 MPa m 1/z and simultaneously exposed to the water environment during irradi-
J. Fohl et al. / Irradiation experiments in the testing nuclear power plant V A K
ation. Macroscopic examination did not reveal any stress corrosion cracking. The fracture toughness derived from subsequent testing fall within the scatter band of normal K~c testing according to A S T M E 399. An irradiation experiment with simultaneous radiation, and cyclic loading in the water environment is being prepared to be carried out in the VAK.
7. Acknowledgement This investigation was conducted within the research programme "Integrity of Components" (F KS) financed to 50% each by the Ministry of Research and Development (BMFT) and an "Industrial Consortium". A.Gerscha, G. Hofmann, E. HOppner, E. Klausnitzer, J. Koban, W. Schweighofer, H.J. Strobel, G. Vasoukis from K W U Erlangen, have contributed to the planning, performance and evaluation of the experiment. The authors wish to express their gratitude to the institutions that gave financial support, to the participants involved in this experiment, and to the Rheinisch Westf~ilische Elektrizit~itswerk (RWE), the
79
Bayernwerk A G (BAG), and to the directorate of the Versuchsatomkraftwerk Kahl (VAK) for giving permission and support to perform this irradiation experiment.
References [1] K. Kussmaul, Die Gewahrleistung der Umschliessung, Grundlagen und Nachweis der Berstsicherheit von Reaktordruckbeh~iltern for LWR-Kernkraftwerke, Atomwirtschaft (Juli/Aug. 1978) S. 354-361. [2] J. F6hl et al., Bestrahlungsuntersuchungen im Forschungsvorhaben Komponentensicherheit, 5. MPA-Seminar 1979, Sicherheit der druckftihrenden Umschliessung von Leichtwasserreaktoren, Stuttgart 11.u.12. Okt. 1979. [3] E. Klausnitzer et al., Irradiation behavior of nickel-chromium-molybdenum type weld metal, ASTM/STP 683, Ninth Intern. Symp. on Effects of Radiation Embrittlement on Structural Materials, 1978 Richland, Wash. USA. [4] W.N. Mc Elroy et al., LWR-pressure vessel surveillance dosimetry improvement program 1979, Annual Report N U R E G / C R 1291. [5] F.J. Loss et al., Fatigue crack growth rates of irradiated pressure vessel steels in simulated nuclear coolant enviroment, J. Nucl. Mater. 96 (1981) 261-268.
65
IRRADIATION E X P E R I M E N T S IN THE TESTING NUCLEAR POWER PLANT VAK J. F O H L
Staatliche Materialpriifungsanstalt (MPA), University of Stuttgart, Stuttgart, IV. Germany Ch. L E I T Z
Kraftwerk Union AG (KWU), Erlangen, W. Germany and D. A N D E R S
Forschungszentrum Geesthacht (GKSS), IV. Germam' Received 19 March 1982 Within the German Research Programme "Integrity of Components" the first two capsules were irradiated in the Testing Nuclear Power Reactor VAK. The materials are of the 22 NiMoCr 3 7 and 20 MnMoNi 5 5 types and represent the lower bound of the base material regarding upper shelf energy and chemical composition (Cu, S, P), as well as a state of material which does not meet both chemical and toughness requirements (low upper shelf test melt). Tensile, Charpy, drop-weight, and fracture mechanics specimens were irradiated up to a range of 1.5 to 2 × 10t9 cm 2 ( E >1 MeV). Despite the materials being at or beyond the specification limits, the results show irradiation sensitivity which can be predicted from the US Reg. Guide Trend Curves (1.99) and KWU Trend Curves in a conservative manner. The procedure to determine the adjusted reference temperature RTND-r (adj.) on the basis of AT41J (following ASTM E 185) could also be confirmed as conservative by comparing the different criteria derived from Charpy and drop weight tests in the unirradiated and irradiated condition. The results of fracture mechanics testing in the linear elastic range show a remarkable temperature margin to the Kl:curve of ASME XI. Prestrained compact tension specimens CT 40 mm made of 22 NiMoCr 3 7 material with an upper shelf energy of approx. 100 J were wedge loaded in a range up to 30 MPa m 1 / 2 and exposed to the water environment during radiation. Macroscopic examination gave no indications of stress corrosion cracking. From tests of these specimens in the linear elastic range, a fracture toughness K~'~, which was not affected by the prestrain and environment history, was found depending only on the overload applied during the prestraining procedure.
1. Objective of investigation N e u t r o n radiation causes a change in the mechanical properites of materials. The influence expected during the complete time of operation according to the conservative trend curves is taken into account for the design calculation of reactor pressure vessels. With the help of surveillance specimens which are irradiated in a position of higher flux than at the inner surface of the pressure vessel wall, the actual change in properties is experimentally determined so that the conservativism of the assumption can be confirmed. For new plants built according to the basis safety concept [1], only a little change in toughness is to be expected. This is based on the present trend curves due to the restricted chemical composition and the limita002%5493/82/0000-0000/$02.75
tion of fluence by 1 × 1019 cm -2. Therefore, the surveillance programmes of such plants include only a minim u m number of Charpy and tensile specimens to validate sufficient material toughness for the whole time of operation. Vessels of which the material has higher sensitivity against neutron embrittlement due to chemical composition and which are exposed to a higher neutron fluence cannot be valuated alone on the basis of Charpy energy in case of a low toughness level. Each single case requires a complete fracture mechanics analysis. The radiation of the necessary large specimens in the respective power reactor is usually not possible. The research p r o g r a m m e "Integrity of C o m p o n e n t s " (F KS) is focussed on the following tasks - the lower toughness level at which safe operation
© 1982 N o r t h - H o l l a n d
66
J. F6hl et al. / Irradiation experiments in the testing nuclear power plant VAK
can still be assured, how to transfer results from small scale specimen testing to the c o m p o n e n t behaviour, - which criteria describe the material toughness and whether the behaviour of low toughness material achieved by metallurgical and thermo-mechanical processes represent the behaviour of neutron embrittled material. The aim of the irradiation experiments of the F KS is to determine fracture mechanics properties of irradiated materials and to establish the limit of possible irradiation embrittlement even for extreme material states in the as delivered conditions. In order to obtain results in a short period of time, most of the irradiation experiments are carried out at high neutron flux in research reactors, since large specimens can be irradiated only there. Such short-time irradiation experiments involve additional parameters, such as neutron flux and neutron energy spectrum. Therefore, further corrections are necessary for applying the results to the pressure vessel wall [2]. The differences in neutron spectra for a variety of reactors and surveillance positions are shown in fig. 1, representing the influence of neutron spectra on the primary damage rate (displacement per atom = dpa). The damage rate by neutrons with an energy E > 1 MeV in the capsule of the research reactor F R G - 2 (CK 120 type capsule b e h i n d a steel shield) a m o u n t s to approx. 30%, in the V A K radiation position approx. 55%, and at the RPV wall (inner surface) of a 1300 M W power reactor approx. 80%. The main portion of material damage is included, if only the damage caused by neutrons in the energy range E > I MeV of a power reactor is calculated. However, this portion is only
power reactors ~ > lO0
i
~
~
~
""
b
-
% 50
0
neutron energy
~
=
a
b
b
~--
o
b
~_~IMeV
• [] ~ []
E~3MeV 3~-E>1MeV 1~-E~0,1MeV E~-O.1MeV
o =surveiU, pos. b : RPV inner surface
z so o ~
~_
_
0
Fig. 1. Comparison of neutron spectra of research and power reactors.
approx. 1 / 3 in a research reactor. Therefore, the effective change in material properties is greater after irradiation in the research reactor than at a comparable fluence E > 1 MeV in a power reactor. The comparison of the spectra a m o n g power reactors at the pressure vessel wall shows that the differences are not so severe. There are differences of about three orders between n e u t r o n flux at the RPV-wall of a 1300 M W reactor and the irradiation position in the FRG-2. The current available knowledge about the effect of dose rate on the material embrittlement is not yet sufficient. However, it can be assumed that high flux radiation experiments cover the embrittlement of reactor pressure vessels in a conservative way, since the time is shorter for recovery during radiation. The F KS irradiation p r o g r a m m e includes a great n u m b e r of various irradiation positions in different reactors as well as a wide range of materials from which those single parameters can gradually be assessed. The irradiation experiments conducted in the Testing Nuclear Power Reactor Kahl (VAK) are an i m p o r t a n t link in this chain of transferability of results from research reactors to surveillance programmes, and finally, to assess the safety of the reactor pressure vessel. Furthermore, it is possible to investigate the superposition of parameters during radiation as they are cyclic loading, water environment, temperature and neutron radiation. This paper presents the results from the first F KS irradiation experiment carried out in the VAK.
2.
Materials
For the first VAK-irradiation experiment, the following materials were selected: - KS 01 forged shell 22 N i M o C r 37, fig. 2; lower limit of specified upper shelf energy (80-100 J). KS 12 forged shell 20 M n M o N i 55, fig. 3, manufactured after M H K W * procedure at the b o u n d of the specified chemical composition. KS 07 B forged slab, fig. 4, with chemical composition exceeding the specified limits of 22 N i M o C r 37 "Low Shelf Test M e l t " (LSTM) with an upper shelf energy of 3 5 - 4 0 J. T h e Charpy energy curves of the materials investigated are represented in fig. 5 together with material as currently used in commercial plants (KS 13). * In the MHKW process a trepan is removed from the ingot and the core is subsequently refilled by electroslag remelting (ESR).
C
Si
Mn
P
Chemical composition (wt %) S
~
KS07
0.18 2Q0.28
1.27
148
1.15 1.55
0.62
0.71
) belt line region.
032
p-~
Ks12
KS 13
a guiding value, (
0.10 0.35
0.29
0.24
0.15 0.25
20 MnMoNi 55 Optimized
~
KS01
~-~
<~0.012
[ ~
~
0.007 0.007
~
~<0.012
~
~
22 NiMoCr 37 Specification 0.17 0.50 ~<0.025 ~<0.025 after VdTUV 0.25 ~0.35 1.00 365 (4.72) Optimized ~<0.20 0.20 a 0.85 a ~<0.008 ~<0.008
Material
Table 1 Materials of the 1. VAK irradiation experiment
0.30 0.50
Ni
0.50 0.60 0.80 1.00
Mo
0.04
015
~<0.20
0.49
0.41
0.45 0.72
FCZiq063
0.40 0.45 0.55 0.85
11.-~0.74
['~0.95
<~0.40 ~<0.55 1.20 a
Cr
0.046
0.01 0.04
~<0.05
AI
0.05
~
~<0.12 (~<0.10)
002
~<0.02
0.05
0.012
<~0.01
<~0.05
V
0.022 <0.01
0.027
0.010 0.040
10.---~ ~<0.003
[~]
~<0.10
~<0.20
Cu
N
As
0.007
~<0.011
0.012
0.01
~<0.013
-
-
<0.05
<0.01
~<0.005
Sb
-
Co
<0.005
0.007
0.011
~<0.030 ~<0.030
0.16
_
~<0.030 ~<0.030
Ta
0018 <0.005 <00,
~<0.025
0.026
0.017
<~0.010 ~<0.013 ~<0.015
Sn
~
~
~.
5
~"
5"
~"
t,,
z2
68
J. Fehl et al. / Irradiation experiments in the testing nuclear power plant VAK
-,~ 550
6000
_
KS 01
K S 07 B
22 NiMoCr37
22 NiMoCr 31
/
- mid -ptane
.f,t.b
/q/I
heod
0 , ~S
Fig. 2. Forged shell KS 01, upper shelf energy 80-100 J.
--
LJ/
Fig. 4. Forged slab KS 07 B, low shelf test melt (LSTM) with upper shelf energy of 30-40 J.
KS 12 20 MnMoNi55 ~
318o
ESR
J
unirrodieted
,
~,"
so
-200
/ I- .....
:
l~/
l~
t
-I00
,'
't
+IL-TI "°°' ~KsoTB
.._ .........
0
100
2_
KS NDT-T "C 01 +5 07B (*30) f2 con~ +10 12 ESR -40 13 -20
200
_/.~
300 °C 1,00
temperoture
Fig. 5. Range of material toughness selected for the irradiation experiments in the research program "Integrity of Components" (F KS).
Fig. 3. Forged shell KS 12 after M H K W procedure, lower bound of chemical composition.
Table 2 Tensile properties of the unirradiated materials
specification
•p0,2 N/ram2
R
FIT
350 ° C
RT
350 ° C
RT
390
343
560 70O
505
19
16
45
35
438
379
586
552
25
20
55
45
m
N/ram 2
z%
A5% 350 ° C
RT
350 ° C
22 N i M o C r 37 KS 01 K S 07 B
i1 21 5481
[13
8,3]
20 M n M o N i 55 K S 12 c o n v °
535
433
694
604
20
15
56
46
627
23
21
67
62
550
27
20
73
65
KS 12 E S R
558
436
705
K S 13
437
375
577
69
J. FOhl et al. / Irradiation experiments in the testing nuclear power plant VAK
The chemical composition and the tensile properties are compiled in tables 1 and 2, respectively, together with the specified data of steel 22 N i M o C r 37 (ASTM A 508 C1 2) and 20 M n M o N i 55 (ASTM 533 B C1 1, 508 CI 3). All materials were used in the quenched and tempered condition and simulated post weld heat treated. It has to be taken into consideration that in case of KS 12 it was not possible to remove specimens from the usual quarter thickness location, since the ESR zone has only a thickness of 60 to 80 mm. Therefore, the specimens represent the material state of a highly quenched surface region.
Table 3b Specimens of the 1. VAK irradiation experiment, capsule B1
I K S 01
3.1. Specimen and capsule preparation
The investigation including preparation of specimens and capsules as well as testing in the hot cells was carried out by KWU-Erlangen. Type and number of specimens for the first irradiation experiment are listed in tables 3a and b. Wedge loaded compact tension specimens from KS 01 material were exposed to the water environment to find out if simultaneous radiation will cause stress corrosion cracking. In order to separate effects caused by prestrain from those related to the water environment, wedge loaded specimens were also placed inside the capsule. Due to the geometrical conditions at the capsule position it was only possible to apply compact tension specimens with a thickness of 40 mm (CT 40). Three different stress intensity levels, namely, apTable 3a Specimens of the 1. VAK irradiation experiment, capsule AI capsule
KS 12 conv.reglon KS 12 ESR-region KS 07 c o m o a r i son m,~ter i a l 1)
I = low
m
A 1
Drop W e i g h t P2
Charpy ISO-V 1) fluence
I
f l u e n c e 1)
I
h
I
m
h
18
6
6
6
18 18
h 32)
8
K S 12 cony, region
8
K S 12 ESR-region KS 07 comparison
r~q ]
8
9
6
B
8
16
material
m -medium2
-
10
19
cm
-2 h
=
high
not emcapsulated
3. Irradiation experiment
KS 01
OT 40
f l u e n c e 1)
I = low
material
Oharpy
tensile
material
18
3.2. Irradiation o f the specimens
18 18
m = medium 2 " 1019cm -2
prox. 10, 20, and 30 MPa m I/z, were chosen. To apply the stress by inserting the wedge, the specimens were overloaded by 60 to 1()0% because of the compliance of the specimen-wedge system. Prior to installing the specimens in the reactor, they were exposed to the water environment at 280°C in an autoclave for 14 days so that a magnetite layer could be developed to protect the specimens against corrosion in the cold reactor water from the time of installation until the final heat up to operational temperature. Subsequent loading of the specimens in a tensile machine to achieve partial unloading did not indicate relaxation during the exposure in the autoclave. Two thin walled capsules ("collapsed cans") were used so that good heat transfer from the specimens to the coolant was assured. The arrangement of the specimens in the capsules (AI and B1) is shown in figs. 6 and 7, [2]. For the assembly plan of the capsule A1, the axial flux gradient was taken into account by gradually displacing the specimens towards the core. The corrosion specimens were in a cage through which the water had access to the prestrained specimens. Monitor plates were used in three axial planes to determine the neutron fluence and the temperature during radiation. Each monitor plate contained nine sets of fluence monitors (Fe and Nb), and five sets of melting cones covering the temperature range from 271 to 302°C, fig. 8. Moreover, additional single monitors were placed at the top and the bottom of the capsule.
h = higlq
The Testing Nuclear Power Plant in Kahl is of a boiling water reactor type with a power capacity of 16 MW. At the core barrel, there are three positions at
70
J. F6hl eta/.
/
Irradiation experiments in the testing nuclear power plant V A K
radiation direction
radiation direction
Fig. 6. F KS irradiation experiment in the Testing Nuclear Power Plant Kahl (VAK), capsule AI.
which capsules of the size up to 100 X 150 x 950 m m can be irradiated, fig. 9. A relatively high y-radiation leading to a heat up of the specimens is effective at this location. The reactor was operated at 12 MW, because it is k n o w n that reducing the power to 12 MW, a n d thus, lowering the operation and water temperature will set a n upper temperature limit within the specimen package of 290°C. The actual n e u t r o n flux at this position a m o u n t s to 2.2X 10 ]2 cm - 2 s ] in the centre of the capsule. The n e u t r o n energy distribution can be seen in fig. 1. Approximately 15% of the neutrons have an energy above 1 MeV, however, they cause 55% of the primary damage. In comparison with the spectrum of F R G - 2 , where only 30% of the primary damage is caused by neutrons with an energy above 1 MeV, the V A K spectrum is harder, but not so h a r d as the spectrum of a 1300 M W reactor at the vessel wall. The water conditions i m p o r t a n t to the specimens not encapsulated are compiled in table 4. Both capsules were radiated in the V A K at the same
.
_
Fig. 7. F KS irradiation experiment in the Testing Nuclear Power Plant Kahl (VAK), capsule BI.
time during a period of 100 full power days to achieve the desired fluence of 2 X 1019 c m 2 in the capsule mid-plane.
VAK irradiatioh experiment monitor plate
p0e I"~ ~
temperature monitors.
""~" :
271
zs~ 286
~, i @
÷@@
8
i ~9
288 289
295 ~o2
~
i -
# ;
L
L :
~e e • .@
1~6,5
Fig, 8. Monitor plate with temperature and fluence monitors.
J. F6hl et al. / Irradiation experiments in the testing nuclear power plant VAK
71
10zo cm-2 6 3r
tu 2
-6-
o~ 10~g 8 = ti
capsule ~
1i~*
c
/
capsule Bt
2 101°
Fig. 9, Irradiation positions in the Testing Nuclear Power Plant Kahl (VAK).
0
50 mm distance from capsule inner surface
100
Fig. 10. Comparison of fluence monitors evaluation with calculated data. Table 4 Composition of the VAK cooling water
Oxygen
0,9
50
Chloride
3,5 ppm ppb
pH - v a l u e
5,5
-
6
el. conductivity
17
-
71
HS/cm
4. Evaluation of the irradiation experiment
absolute fluence in axial direction of tile middle axis was derived from all iron detectorsl According to the theoretical gradient in tangential a n d circumferential direction, all fluence data of one plane were transformed to the mid-point of the m o n i t o r plate. F o r each mid-point a mean was calculated using a least square fit. All mid-point fluences were fitted by a smooth curve in axial direction, fig. 11. This was the basis from which the fluence of each specimen could be calculated. The reference point of the fluence of each specimen was chosen as the mid-point of the tensile, C h a r p y and drop-weight specimens a n d as the m e a n crack length ( a 2 4 - a 3 + a 4 ) / 3 at B / 2 position of the compact tension specimens.
4.1. Fluence 4.2. Irradiation temperature The n e u t r o n field at the core mid-plane was d e t e r m i n e d o n the basis of the actual b u r n up state of fuel elements. This was c o n d u c t e d with the two-dimensional transport code DOT-3 using the n e u t r o n crosssection data of E U R O L I B 4 with 53 energy groups. The calculated fluence a n d the results of the m o n i t o r evaluation of iron (54 Fe (n, p) 54 Mn), a n d N i o b i u m (93 N b (n, n') 93 m N b ) are c o m p a r e d in fig. 10. The calculated values are conservative since the radial gradient of the power distribution of the edge fuel elements was not taken into account. The agreement between the fluence data derived from N b a n d Fe is good (15%) near to the core side of the capsule, however, at the rear of the capsule, the evaluation of N b leads to a 50% lower fluence. The
F r o m the melting cones of the six m o n i t o r plates an irradiation tempe~'ature was evaluated by visual examination. In all cases, the 288°C m o n i t o r was not melted but the 271 °C one. A differentiated e x a m i n a t i o n of the m o n i t o r s with melting points of 284°C and 286°C led to an irradiation temperature in the range of 282°C to 288°C. 4. 3. Testing o f irradiated specimens 4. 3,1. Tensile tests The specimens were tested in the unirradiated condition in the t e m p e r a t u r e range from - I00 up to 350°C, a n d in the irradiated condition from - 150 up to 275°C.
72
J. F6hl et al. / Irradiation experiments in the testing nuclear power plant VAK
500~-! mm~- .... I,O0 - -~
ii
~0o
i E
i
200
"5
"° 100
L
° 0,
]
"-
I
I
i
°E O.5.1019crn "2 t5 ~c~ 100f|ue nce!(Fe)~
~ 2
2oo~ x o
I 4 I I L
; I
i t
2L5/
!
I
Itt
I '
III
300
I I I Ill
~00
v
500
7
Fig. 11. Fluence distribution in the capsules in axial, tangential and radial direction.
The yield strength, as a function of temperature was established to evaluate the validity of fracture toughness. General information concerning the change in tensile properties due to irradiation can be achieved by c o m p a r i n g results from testing at room temperature a n d testing at 275°C. It has to be m e n t i o n e d that irradiated specimens were not tested at room temperature but at 0°C, fig. 12. However, this does not affect the comparison, since the change in properties in this temperature range is neglegible a n d c a n n o t even be represented in
~ unirracliatecl
~" 1200 r .£~ Nlmm21
1 RT J
800 1~
the figure. U l t i m a t e and yield strength are slightly higher in the irradiated condition than in the unirradiated one. Reduction of area a n d elongation, which are usually lower after irradiation show, do not show in every case this tendency. The change in these properties is small a n d the differences can p r o b a b l y be traced back to material a n d test scatter. 4.3.2. Charpy impact tests The test temperatures were chosen in a range up to
~ irradinted VAK 1,5-1,8.10 ~9crn2 ~)tested ot 0 °C
275 °C
O:E1201 ~N/mm
RT 11
275 °C
-
"~, KS 01 30
I
KS12 conv. ESR
KS 01
RT 1)
100
,E
.9 E 0
KS 12 cony, ESE
RT 11
%
E KS 01
KS12 cony. ESR
o KS 01
Fig. 12. Change in tensile properties due to neutron irradiation.
KS 12 cony. ESR
275 =C
1
[
J. Fohl et aL
/
Irradiation experiments in the testing nuclear power plant V.4K
275°C in the irradiated condition so that a complete energy temperature curve could be developed including low shelf, transition region and upper shelf. One specimen was usually tested at each temperature step but in the range of the assumed energy of 68 J three specimens were tested at one temperature. The Charpy energy curves for the unirradiated and irradiated specimens of one fluence level are plotted in figs. 13 through 16. The figure shows also the temperature printed at which the energy of 68 J and 41 J, respectively, and the lateral expansion of 0,9 mm, usually the characterizing criteria, is reached. All criteria before and after irradiation are compiled in table 5. The shift in the transition temperature based on 41 J is compared with the trend curves of US Reg. Guide 1.99, fig. 17. All experimental data fall far below the trend curve corresponding to the phosphorous and copper content. The trend curve of KS 07 B material is identical with the upper limit curve. The trend curves assessed by K W U indicate in some regions a lower transition shift than the Reg. Guide curves. They still represent a conservative shift compared to the experimental results, fig. 18. 4.3.3. Drop weight tests O n l y 6 s p e c i m e n s were a v a i l a b l e to d e t e r m i n e t h e d r o p - w e i g h t N D T - t e m p e r a t u r e . T h e r e f o r e , it w a s n o t p o s s i b l e to e v a l u a t e in e v e r y c a s e a p r o p e r l y d e f i n e d transition temperature. The highest temperature (within a r a n g e of 15 K as t h e g r e a t e s t difference), w h i c h c o u l d b e r e f e r r e d to t w o " n o b r e a k " s p e c i m e n s , w a s t a k e n to e s t a b l i s h t h e d r o p w e i g h t N D T . T h e r e s u l t s are also listed in t a b l e 5. T o r e n d e r p o s s i b l e t h e c o m p a r i s o n o f N D T a n d C h a r p y e n e r g y shift at t h e s a m e fluence, t h e
criteria [*C] unirradinted irradiated 200
J
T6ej /,8 106
1"~1j 25
~
i
KS 01, (T_L) i ~ 22 HiMoCr 37
]
68
criteria [*C] unirrodiated irradiated
20 lOB
F q ~
200
J
°~16o
62
T~9,.. 26 111
(L-T
KS0~
-
i
~100 --
-unirr0dioted
&lJ -200
•
~ -I00
a
o
o
! 1,1jO19crn"2 IE~MeVl /
o 0
i 200 temperature I00
J 300 *C /,00
Fig. 14. Charpy-energy of KS 07 B material as a function of temperature in the unirradiated and irradiated condition•
criteria unirradiated
[*el Tsej 30
KS 12 conventional
20MnMoNi 56
t~160
'
.~ O
~ -- ~
-
~
75
• (T-LI
~
Tos~ 10 103
S~ ~T,---- - ~ ] &l J
7
•
J
~
i
unirr0diat
B8J
-200
T~,j 6
113
irradiated
200 j
[,
----JtZ ~/ irrQdlotedVAK| ~ ~ 1~10~scm2{E>]MeV)
-100
0
IO0 200 temperature
300
*C /.00
Fig. 15. Charpy-energy of KS 12 material (conventional zone) as a function of temperature in the unirradiated and irradiated condition.
criteria [*C] T~ej unirradiated -10 irradiated 90
To~mm 25 77
T~. -32 58
To~mm -1~ 81
200 KS12'ESR (T-L) ~ l t = ~ r . ~ J 20 HnMoNi55
I
;~ 100 I
r~ 6 0 , _ - - ~ _ _ _
T~lj /,
T6aj
73
unirrodi(lted
5
~_---/~eirrQdiQtJdVAKi-~t =-vL~ 4
u
irr(ldi~ted VAK[ 2,0.101Scm"z (E>IM I
I
-200 -200
-I00
0
I00 200 temperature
300
*C i,O0
Fig. 13. Charpy-energy of KS 01 material as a funktion of temperature in the unirradiated and irradiated condition.
-100
0
100
temperoture
200
300 *C tOO
Fig. 16. Charpy-energy of KS 12 material (ESR zone) as a function of temperature in the unirradiated and irradiated condition.
unirrad. irrad.
unirrad. irrad.
unirrad. irrad.
KS 12 konv.
KS 12 ESR
KS 07
adjusted fluence. b after [4].
unirrad. irrad. irrad.
KS 01
Material
1,1 X 1019
X
1019 1,6X 1019
2
25 111
81
-- 14
--
20 108
90
-- 10
136
4 62
32 58
100
40
--40
75
121
75
3,0× 1019 2,5 X 1019
113 65
10
45
5
103
5 75
25 68
NDT
1,9X 1019 1,5X 1019
30 110
48 106
T41 j
55
10 93
25 77
T68 j
1,1X 1019 0,9X 1019
2,1X 1019 1,7 X 1019
To.9 mm
Temperature (°C)
Table 5 Comparison of criteria to describe neutron embrittlement
86
95
111
93
83
52
ATo.9 mm
88
100
106
83
80
58
AT68 J
Temperature shift
58
90
95
70
70
43
AT41 j
t,v90 a 80
if70 a \ 65
ff62 a \ 55
if50 ~ 45
F45 ~ \ 4O
ANDT
119
122
136
121
103
67
chemistry relation
AT b
5
e~
t~
2
75
J. F6hl et al. / Irradiation experiments in the testing nuclear power plant VAK
/,00 irrod' temp. 2'08'°C ' after 00'Reg'.Guide 1.99i77 K ---trend curve KSQ1 • VAK-exp.(AT61J} .~. 200 . . . . . . KS 12 cony. * " ~ , a p ~ KS 12 ESR o " ~ ~]~."~ -~S ~ ' /
~u
100 80
~
so
~L_r.~ --
~
t --~-~"
I I
. . . . . . Cu .
• KS 01 K -* KS 12 cony. <1 o KS 12 ESR ~: D KS 078
I
P
.11 .009 17 .015 .26 .022
I I e
E
~
I 0u :0,10 % r p =0,012Ol,
~35
t2ol I
I
o,lo'/.ou]
[
I '/.
3,oo
g
50 -
10
T-L ,, ,, L-T
+ °A~-'~
~ ~o ~ 20 -,~
150
]
. . . . . . . . . .
2x10 ~7
6
8 10~e 2 ftuence
711 t, 6 ( E > I 1,4eV)
t.
I I'l
2
I 3
-
..... 10TM
2
cm -2
0 ~IP=0:012, 10le 2
3
"/*, a,fterKWU t, 5 6 7 8910 TM
|
i J/ t, 5 0 cm"z 10 z°
neutron fluence e (E>IMeV)
Fig. 17. Comparison of transition temperature shift (T4~j) with the trend curves of US Reg. Guide 1.99.
N D T shift was extrapolated on the basis of the slope of the Reg. G u i d e 1.99 trend curves (AT--q~1/2). The results show that the shift in transition temperature measured at a level of 41 J according to A S T M E 185 a n d K A T draft 3203 is almost equal to the drop weight N D T shift of KS 01 and KS 12 ESR material, b u t greater for KS 12 conventional material. The energy criteria for the u n i r r a d i a t e d specimens a n d the shift caused by n e u t r o n irradiation are plotted in fig. 19. T h e adjusted reference temperature RTND T (adj.) of the irradiated material (closed triangle), derived from the reference t e m p e r a t u r e R T N o T* according to A S M E III (open triangle) a n d the t e m p e r a t u r e shift at the 41 J level, is in all cases higher than the drop weight N D T - t e m p e r a t u r e of irradiated specimens. The procedure based on AT41j following A S T M E 185 to determine the reference t e m p e r a t u r e of irradiated materials could be confirmed as conservative. This was already d e m o n s t r a t e d in previous investigations with high copper weld material [3]. The transition t e m p e r a t u r e shift derived from the chemical c o m p o s i t o n (chemistry relation [4]) covers all experimental data d e t e r m i n e d at 41 J in a very conservative way. 4.3.4. Fracture mechanics tests The fracture mechanics specimens were fatigue precracked in the u n i r r a d i a t e d state. The direction of radi-
* The NTNDT was determined in this case from the Charpy energy mean curve. ASME requires the lowest value of three specimens tested at N D T + 3 3 K being at least 68 J for the energy and 0.9 mm for the lateral expansion.
Fig. 18. Comparison of transition temperature shift (T41J) with the KWU-design curves.
ation was so that the u n a v o i d a b l e fluence gradient occurred in the direction of the crack extension, cf. figs. 7 a n d 11. T h e specimens were tested in a temperature range where valid Kit data after A S T M E 399 could be assumed. The fluence of the compact tension specimens a m o u n t e d to approx. 1,5 × 1019 cm -2 ( E > 1 MeV). To verify the validity of fracture toughness, the yield strength was interpolated in the diagram Rp0.2 = f ( T ) . In the u n i r r a d i a t e d condition specimens were tested up to a thickness of I00 m m (KS 01 a n d KS 12 conv.). As the ESR zone of K 12 is limited in thickness direction, it is not possible to use a larger specimen than C T 40 (cf. section 2). Therefore, Kit-testing was re-
material KS 01
fluence
-50
0
temperature 50
100
•
150
2,1.1011¢m-2JE~IHeVIj ~
m.--.*.----.* IlOl KS 12 1,1.1019 conventioneL
I
1,.9.10z9
!
I
3.0.10t9
KS 12 ESR
2,0,10TM
KS O7 B
1,1.10 ly
Fig. 19. Comparison of criteria derived from Charpy impact testing with the NDT-temperature before and after irradiation; the RTNDT was determined from the Charpy-energy/temperatare mean curve.
76
J. FOhl et al. / Irradiation experiments in the testing nuclear power plant VA K
stricted to only a small temperature range. The test results of the unirradiated and irradiated specimens are plotted in figs. 20 through 22. As only a limited number of specimens were available and because of the relatively large scatter, it is not possible to draw a curve through the measure points. The Klc-CUrve and the adjusted Kit-curve (according to A S M E XI) corresponding to R T N o T and RTNt)T (adj.) are given as reference in the plots. The Ki~-range of the conventional zone of KS 12 determined with CT 40 specimens at a fluence of 1.5 × 1019 cm 2 extend almost up to room temperature. The conservatism of the Kt~-curve of A S M E XI could be demonstrated for all three materials in the unirradiated as well as in the irradiated state. The temperature difference between the experimental results and the KI~ curve of A S M E XI amounts to approx. 40 K and more. 4. 3. 5. Testing of prestrained specimens The visual examination of the wedge loaded specimens indicated no crack growth. The specimens were loaded in a tensile machine and the l o a d / d i s p l a c e m e n t curves plotted. The specimens were overloaded up to 70-200% in order to remove the wedges. F r o m the load displacement curves, the remaining stress intensity was determined. It decreased by approx, up to 14% due to relaxation. There was no difference between specimens in the capsule and outside the capsule. After removing the wedges all specimens were tested at - 5 0 ° C and a K ~ or KQ-Value, respectively, was determined, table6. Macroscopic examination of the fracture surface at the crack tip region did not reveal stress corrosion cracking. The Kjc and the KQ values,
KS 01 NiMoCr 37 o unirradinted • irradiated VAK 1,5.101~crnI[E-1 MeV}
22
M Pa
u ' ~ 100
/~
==
~ ZXT~Ij
2 o
=
50~
i
oL___ -200
KIe after ASME Xl / j~unirradiated . adjusted (ATt.ll)
4000 N mm"3n
3000
j
J ladjl 200
---J~adjusted
&O00
Nmm"If;
{AT{,,j)
'
3000
!
2000
RTWO I
RINDT4udjl
i
-200
_
-I00
0
I00 temperature
200
J o
*C
300
Fig. 21. Fracture toughness (linear elastic) of KS 12 material (conventional zone) before and after irradiation in comparison with the Kmccurve of ASME XI.
respectively, as a function of effective prestrain, are shown in fig. 23 (left). Kmc could not be determined with one specimen, since the fatigue crack propagated only on the surface, but not in the centre of the specimen. Ktc and KQ, respectively, show an ascending tendency with increasing prestrain (prestress). More important than the stress intensity during irradiation is probably the maximum load applied during prestraining, which causes blunting of the fatigue crack, fig. 23 (center). The fracture toughness is not influenced up to a Km value of approx. 35 MPa m 1/2 (1100 N m m 3/2). The fracture toughness of the specimens irradiated regularly without prestrain is again shown in fig. 23 (right), (ef. fig. 20),
MPa~rm
:~
100
KS ~2 ESR 20 MnMoNi 55 o unirradioted • irradiated VAK
1,5I0~9cm ~ (E>IMeV) 1o)
c
1000
100 temperoture
Klcafter ASM£ Xl
o
= 5o!: U o RTNot
!
:
j/~ unirradiated
lOOO
I
•
~
/
!
2000
oo o
-100
goO
(*1
i io
I
F
150~
T
150
ISO F - T ! KS 12 conv. 20 MnNoNi 55 M P u ~ ] ounirrudinted • irradiated VAK 100 1'5 101' cm ~ (E'IMeV)
=C
J 0 300
Fig. 20. Fracture toughness (linear elastic) of KS 01 material before and after irradiation in comparison with the Kiccurve of ASME XI.
0i__L -200
-100
•
~ I/T~/~
Kicoffer ASME XI . unirradiated adjusted (Alto))
&O00 N mm
3~;
3000
2000
1000
,T,,o, ,ooj, 0
0
I00 temperature
200
°C
300
Fig. 22. Fracture toughness (linear elastic) of KS 12 material (ESR zone) before and after irradiation in comparison with the Kk.-curve of ASME XI.
77
J. ['Oh/et al. / Irradiation experiments in the testing nuclear power plant V A K
KS 01 22 IfiHoCr 37 test temperature -50 or • in c0psule [-irradioted 1,5.10~g cm-2 • in water Lwith prestrain
150
150
MPo~
~g
~ lOO
irrodioted 1.5 -10~9cm"2
l
i
HP,
e
I00
c
2
5o
i"
.ronge with
l J
overtoed
50 , e ~ I
o
,
25 50 MPo~mm K1 [prestroin)
0
25 50 MPolmm K] (overlood)
-100
"C
100
temperature
Fig. 23. Fracture toughness of specimens irradiated with prestrain (with and without exposure to the cooling water).
together with the results of the prestrained ones. It is evident that, in this case, prestrain and water environment have no effect on fracture toughness.
5. Cyclic loading test The knowledge of crack growth behaviour is necessary for the lay out of reactor components. It is a well established fact that crack growth rate under cyclic loading is higher in water environment than in air. Some specimens were tested occasionally in an autoclave after irradiation. It can be stated from these tests that irradiation has no effect on cyclic crack growth rate [5]. However, it is not known, whether this statement is also true for simultaneous radiation and cyclic loading as there have not been carried out any tests. It cannot be excluded that the coolant (electrolyte) will change during neutron irradiation in such a way that crack growth will be accelerated due to a higher aggressiveness. In addition to the experiments performed already with static load, a test rig is being built at GKSS. This is capable of applying cyclic load to 40 mm thick compact tension specimens, simultaneously with irradiating in the cooling water environment. The cyclic load is generated by an alternating helium pressure in bellows. The load is directly measured at the end of the specimen chain consisting of 5 CT 40 specimens, fig. 24.
% !% % %
(~) C T 4 0 specimens (~) load bellows ~) coupling links ~)
f
supporting frame
(~) force transmitter (~) t r a c t i o n rods (~) Instrument leeds (~) guard plates
• /
~)
spacer block
Fig. 24. Irradiation rig for crack growth investigations.
78
J. FOhl et al. / Irradiation experiments in the testing nuclear power plant VAK
Each specimen is instrumented with a radiation resistent differential transducer to measure the crack opening displacement. The corresponding crack length can be determined from the compliance curves so that it is possible to evaluate complete d a / d N = f ( A K ) curves. The necessary helium pressure as a function of time is generated in a pressure transducer loaded by a mechanical tensile testing machine• A function generator obtaining its signal from the load cell inside the reactor, controls any load-time function at a frequency of approx. 1 cpm and lower (up to 1 cph). A dummy-run with unloaded specimens will serve as a general check of the instrumentation system and its long-time behaviour. Furthermore, the temperature at the crack tip and in the surrounding water will be measured• The first cyclic experiment is to start at the end of 1982.
tt
.o dl
© t.~ i-
6. Summary E ol
.=_
~r',
< > O
.=_
o£
e~
I.
<<<
<
&
<<<
Within the Research Programme "Integrity of Components" (F KS) two capsules were irradiated in the Testing Nuclear Power Plant Kahl (VAK). The materials KS 01 and KS 07 (22 N i M o C r 37) and KS 12 (20 M n M o N i 55) were irradiated and subsequently tested in the hot cells at K W U . The materials represent lower bound conditions concerning chemical composition (KS 12), and upper shelf Charpy energy (KS 01) and both (KS 07 B), respectively. Tensile, Charpy, drop-weight and compact tension specimens were irradiated. In spite of materials that are at or beyond the specification limit, the results show irradiation sensitivity which can be predicted in a conservative manner from the US Reg. Guide Trend Curves (1.99). The procedure of determining the adjusted reference temperature R T N D v (adj) on the basis of 41 J (following A S T M E 185) was also confirmed as conservative• This has been obtained by comparing the different criteria derived from Charpy and dropweight test in the unirradiated and irradiated condition. It was not possible to determine the shift in fracture toughness, because of the limited number of specimens tested in the unirradiated and irradiated condition and scatter in the test results. The results of fracture mechanics testing show a remarkable temperature margin to the Kl~-curve of A S M E XI. Compact tension specimens of 22 N i M o C r 37 with an upper shelf energy of approx. 100 J were wedge loaded in a range of up to 30 MPa m 1/z and simultaneously exposed to the water environment during irradi-
J. Fohl et al. / Irradiation experiments in the testing nuclear power plant V A K
ation. Macroscopic examination did not reveal any stress corrosion cracking. The fracture toughness derived from subsequent testing fall within the scatter band of normal K~c testing according to A S T M E 399. An irradiation experiment with simultaneous radiation, and cyclic loading in the water environment is being prepared to be carried out in the VAK.
7. Acknowledgement This investigation was conducted within the research programme "Integrity of Components" (F KS) financed to 50% each by the Ministry of Research and Development (BMFT) and an "Industrial Consortium". A.Gerscha, G. Hofmann, E. HOppner, E. Klausnitzer, J. Koban, W. Schweighofer, H.J. Strobel, G. Vasoukis from K W U Erlangen, have contributed to the planning, performance and evaluation of the experiment. The authors wish to express their gratitude to the institutions that gave financial support, to the participants involved in this experiment, and to the Rheinisch Westf~ilische Elektrizit~itswerk (RWE), the
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Bayernwerk A G (BAG), and to the directorate of the Versuchsatomkraftwerk Kahl (VAK) for giving permission and support to perform this irradiation experiment.
References [1] K. Kussmaul, Die Gewahrleistung der Umschliessung, Grundlagen und Nachweis der Berstsicherheit von Reaktordruckbeh~iltern for LWR-Kernkraftwerke, Atomwirtschaft (Juli/Aug. 1978) S. 354-361. [2] J. F6hl et al., Bestrahlungsuntersuchungen im Forschungsvorhaben Komponentensicherheit, 5. MPA-Seminar 1979, Sicherheit der druckftihrenden Umschliessung von Leichtwasserreaktoren, Stuttgart 11.u.12. Okt. 1979. [3] E. Klausnitzer et al., Irradiation behavior of nickel-chromium-molybdenum type weld metal, ASTM/STP 683, Ninth Intern. Symp. on Effects of Radiation Embrittlement on Structural Materials, 1978 Richland, Wash. USA. [4] W.N. Mc Elroy et al., LWR-pressure vessel surveillance dosimetry improvement program 1979, Annual Report N U R E G / C R 1291. [5] F.J. Loss et al., Fatigue crack growth rates of irradiated pressure vessel steels in simulated nuclear coolant enviroment, J. Nucl. Mater. 96 (1981) 261-268.