scc growth behavior of bwr core shroud materials
DESCRIPTION
Scc Growth Behavior of Bwr Core Shroud MaterialsTRANSCRIPT
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and this fact suggests that the degree of hardening assumed in the actual shrouds heat-affected zone (HAZ) should bring little
following.
Recently, it is concerned that the SS of high yieldstrength or hardness tends to provide high SCC growth
acquired to prepare a new diagram for hardened SS [3].Since the reactor core shroud suffers from neutron
of the disposition curve for sensitized SUS304,9.2 1010m/s, is tentatively supposed to be applied incase of uence higher than the threshold. Needless to say,
ARTICLE IN PRESSaccumulation of SCC growth data of the irradiatedmaterial has been anticipated to enable more accurateand reasonable life prediction of the component.
0308-0161/$ - see front matter r 2008 Elsevier Ltd. All rights reserved.
doi:10.1016/j.ijpvp.2008.01.004
Corresponding author.E-mail address: [email protected] (H. Yamashita). Nearly 3601 circumferential discrete cracks existed in thering along the weld between the shroud shell and thering. Most of the cracks were several mm away from thefusion line.
Several radial cracks were found in the heat-affectedzone (HAZ) of H4 weld, the highest uence position inthe shroud, some of which propagated into the weldmetal.
irradiation, it would provide different SCC behaviors.Main factors for increasing crack growth rates seem to bethe increased yield strength due to the cyclic weld heatinput and the neutron irradiation effect.In assessing the SCC growth rates of BWR core shrouds
whose uence is below the IASCC threshold uence,5 1024 n/m2, the K-da/dt disposition curve for un-irradiated SS is applied. And the upper crack growth rater 2008 Elsevier Ltd. All rights reserved.
Keywords: IASCC; Crack growth rate; Core shroud; SCC
1. Introduction
In Japanese BWR plants, stress corrosion cracking(SCC) in components such as the reactor core shroudand primary loop re-circulation (PLR) piping made of L-grade stainless steel (SS) has been an issue since 2003 [1,2].Investigation on SCC in the core shroud has revealed the
rate. The da/dt-K diagrams that were prepared based onL-grade SS (heat treatment with 620 1C 24 h) SCC datamight not envelope the condition of radiation-hardenedand/or cold worked shroud. Incidentally, in the process ofinvestigating PLR piping SCC, the hardness of weld HAZwas found to be around 250 HV and many SCC growthdata with the PLR piping weld joint mock-ups wereenhancement effect in SCC growth rates.Abstract
Recent studies suggest that material hardening should cause enh
steels (SS). In this study, SCC growth rates of SS irradiated up to
obtain the reasonable estimation for SCC growth behavior in the
actual BWR components (a core shroud made of 304SS and a top g
and H4 shroud weld mock-ups made of 316L. In irradiated SS, SCC
to be 1010m/s. In un-irradiated SS, all SCC growth rates were beaUtsue Valve Service Co, ShbTokyo Electric Power Company, Uc
Received 13 February 2007; received in revised fInternational Journal of Pressure Vesse
SCC growth behavior of B
H. Yamashitaa,, S. Ookib, Y. Tanakand Piping 85 (2008) 582592
R core shroud materials
K. Takamorib, K. Asanob, S. Suzukib
achi, Nishiku, Osaka, Japan
iwai-cho, Chiyoda-ku, Tokyo, Japan
23 December 2007; accepted 16 January 2008
ement in stress corrosion cracking (SCC) growth rate of stainless
2 1025 n/m2 and of un-irradiated SS were measured in order tore shrouds, using 0.5, 0.7, 1.0T-CT specimens prepared from the
e made of 316SS) as well as 0.5T-CT specimens prepared from H3
owth rate in actual core shroud of high uence was estimated later
the K-da/dt disposition curve of the JSME NA1-2002 standard,
www.elsevier.com/locate/ijpvp
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As is assumed recently, difculty in determining theappropriate magnitude of K does exist in performing SCCgrowth tests with irradiated materials. Some provisions inASTM, E399, E647 and E813, give a guideline for therelationship among K, yield strength and the specimen size.Since the yield strength of highly irradiated SS signicantlysoars, very high K would be allowable if those provisionswere applied in due form. On the contrary, Andresen [4]
and Janssen [5] have proposed that special considerationshould be necessary to properly estimate the effective yieldstrength for the irradiated SS.In this paper, the crack growth rate of the reactor core
shroud is discussed based on the following experimentaldata and the previous knowledge.
(1) SCC growth rate data with CT specimens made from
0ma
ARTICLE IN PRESS
H1
H2H3
H4
H6aH6bH7a
H7b
RPV
Core Shroud
Shroud Support
Fig. 1. Reactor core shroud [14].
Steel Ingots Hot Rolled
MachiningCutting
BendingMachining
Welding
SHT W.Q. PickleHeating
Shell
Ring
Steel Ingots Hot Rolled
MachiningCutting
BendingMachining
Welding
SHT W.Q. PickleHeating
Shell
Ring
Fig. 2. Preparation procedure of the 316L H3 weld mock-up [14].
) [1
M
0.9
0.8
0.7
H. Yamashita et al. / International Journal of Pressure Vessels and Piping 85 (2008) 582592 583Table 1
Chemical components of the H3 and core shroud mock-up material (wt%
Location Material C Si
H3 ring SUS316L 0.012 0.47
H3 shell SUS316L 0.007 0.44
H3 ller metal Y316L 0.019 0.40H4 shell SUS316L 0.012 0.48 0.8
H4 ller metal WEL316UCL 0.016 0.46 1.34]
n P S Ni Cr Mo
1 0.021 0.002 12.15 16.62 2.16
6 0.023 0.001 12.30 17.55 2.14
5 0.006 0.001 11.53 18.76 2.137
8stant loading by load cell. Water chemistries during the
in Fcon.5T-CT specimens for the crack growth rate test werede from the two mock-ups. The specimens geometry isig. 4. Loading condition of SCC growth rate tests wasH3 and H4 weld in a shroud mock-up.(2) SCC growth rate data with 1.0TCT/0.7TCT specimens
made of the actually used core shroud and 0.5TCTfrom the top guide.
In addition, it is known that welding residual stress willbe relaxed with high neutron uence irradiation. When weevaluate SCC propagation behavior of the core shroud H4HAZ, radiation-induced stress relief should be taken intoconsideration.
2. Experiment
2.1. Un-irradiated material: H3 and H4 weld mock-up
A concern is that yield strength increase caused by theresidual strain due to the weld heat input would acceleratethe SCC growth rate. In order to study this effect, twomock-ups, H3 weld mock-up and H4 weld mock-up,of the reactor core shroud of 1100MWe BWR plant wereprepared, and specimens were made from parts adjacent toand far from the weld fusion lines.Fig. 1 is an entire view of the core shroud. Fig. 2 shows
the preparation procedures of the H3 weld mock-up. Thechemical composition of the material is shown in Table 1.The scale of the H3 weld shroud mock-up is 1/8 in thecircumferential direction, i.e., a 451 semi-circular shape of2100mm circumferential length. Those in the axial direc-tion use 1/1 for the ring and 1/4 for the shell, respectively.Fig. 3 shows the procedures of the H4 weld mock-up.
The chemical composition of the material is shown inTable 1. H4 weld was simulated by forming grooves on atplates and SMAW welding, under a condition identical tothat of the actual 1100MWe BWR plant.0.018 0.006 12.64 16.63 2.05
0.013 0.006 12.22 19.51 2.22
-
tests were simulated NWC condition (288 1C, 2239 ppmDO, ECP +200mV_SHE) and HWC condition (288 1C,100150 ppb DO and 4050 ppb DH, ECP 160170mV_SHE).
2.2. Irradiated materials: actually used BWR core shroud
and top guide
The core shroud (SUS304) and the top guide (SUS316)actually used for about 20 years in BWR plants of Tokyo
Electric Power Company were provided for the investiga-tion on SCC growth rates of the neutron irradiated SS.Their chemical compositions quoted from the mill sheetsare shown in Table 2.Fundamental material properties of the shroud and
top guide were acquired by conducting the followingtests.
Vickers hardness measurement (1.964.9N loading) Tensile test Metallographic observation after 10% oxalic acidetching
X-ray measurement for collection of 3-dimensionalneutron ux distribution.
Large CT specimens made from the shroud and topguide. For an example, the geometry of the 1T-CTspecimen is shown in Fig. 5. 1T-CT specimens were madefrom the base metal of the shroud, 0.7T-CT from HAZ ofthe shroud and 0.5T-CT of 9.1mm thickness from the topguide. Loading condition of SCC growth rate tests wasconstant loading by load cell. Water chemistries during thetests were simulated NWC condition (288 1C, 32 ppm DO)and simulated HWC condition (288 1C, 20 ppb DO or2 ppb DO and 2040 ppb DH).
ARTICLE IN PRESS
EDM Notch
Welding (Inner) Machining Welding (Outer)
500
55
750
Machining1000
Fig. 3. Preparation procedure of the 316L H4 weld mock-up [14].
om
M
0
0
0
H. Yamashita et al. / International Journal of Pressure Vessels and Piping 85 (2008) 582592584Fig. 4. Geometry of 0.5T-CT specimen made from the shroud mock-
ups [14].
Table 2
Chemical components of the actually used core shroud material quoted fr
Material ID Location Materials C Si
1S Shroud base metal SUS304 0.04 0.580.64
2S Shroud base metal SUS304 0.05 0.580.59
3S, 4S, 5S Shroud weld metal SUS304 0.06 0.70T Top guide SUS316 0.059 0.83 1the mill sheets (wt%) [14]
n P S Ni Cr Mo
.900.98 0.0180.020 0.003 9.209.35 18.50
.91 0.021 0.005 9.409.50 18.6518.70
.99 0.025 0.006 9.42 18.58 Fig
shr.62. 5. Geometry of 1T-CT specimen made from the actually used
oud [14].0.030 0.006 11.46 17.40 2.39
-
3. Results
3.1. Un-irradiated material: H3 and H4 weld mock-up
Figs. 6 and 7 show the hardness around weld fusion linesof H3 and H4 weld mock-up. It seems that the hardness
reaches approximately 200230 HV (about 200HV or lessin the ring) in the part close to the fusion line and decreasesaway from it.SCC growth rate test results are shown in Table 3
and Fig. 8. Crack length was monitored with the DCpotential drop during the test and corrected by themicroscope observation on the fracture surfaces of thespecimens after the test. Crack length for crack growthrate calculation was estimated by dividing the crackedarea on the surface by the specimens thickness, sincewelding residual stress was supposed to be released,when specimens were prepared from HAZ of mock-ups. Stress intensity factor, Kapplied, was calculated fromload and specimen geometry in all conditions. As seen inFig. 8, the acquired SCC growth rates are reasonablyenveloped by the disposition curve of JSME NA1-2002standard for L-grade SS in NWC (ECP4150mV_SHE)and HWC (ECPo100mV_SHE), and no signicantdifference between HAZ and sensitized L-grade SS wasfound.
ARTICLE IN PRESS
Fig. 6. Result of micro-vickers hardness measurement (H3 mock-up) [14].
Weld Metal
Base Metal (Shell)
Base Metal(Shell)
182190194
192208206
205209213
212217206
225214214
229222234
210215229
196198203
198208202
206
212
193
206
202 223
198
200
205 234
221
219
218 230
233
219
229 202
189
175
183
203
199
205
197
198226
197
204
213210
222221
228
221216
182190194
192208206
205209213
212217206
225214214
229222234
210215229
196198203
198208202
206
212
193
206
202 223
198
200
205 234
221
219
218 230
233
219
229 202
189
175
183
203
199
205
197
198226
197
204
213210
222221
228
221216
Fig. 7. Result of micro-vickers hardness measurement (H4 mock-up) [14].
Table 3
Condition and result of SCC growth rate test with the H3 and H4 core shrou
Specimen no. Material KappliedMPam0
1.E-10
1.E-09
1.E-08 316L HAZ, DO22-39ppm(NWC)316L Base Metal, DO39ppm(NWC)316L, HAZ, DO100-150ppb, DH40-50ppb (HWC)JSME L-grade SS in NWCJSME L-grade SS in HWC
rack
Gro
wth
Rat
e, m
/s
H. Yamashita et al. / International Journal of Pressure Vessels and Piping 85 (2008) 582592 585CT-H3S H3Ring 2mm from fusion line 30
CT-H3C H3Ring 2mm from fusion line 30
CT-H4 H4Shell 2mm from fusion line 30
CT-H45 H4Shell adjacent to fusion line 20
CT-H42 H4Shell adjacent to fusion line 25
CT-H43 H4Shell 2mm from fusion line 25
CT-H44 H4Shell 100mm from fusion line 25CT-H46 H4Shell adjacent to fusion line 25
CT-S1 H3Shell adjacent to fusion line 25d mock-up [14]
.5 Environment da/dt (m/s)
NWC 288 1C, DO2228 ppm 2.7E11NWC 288 1C, DO2228 ppm 3.1E11NWC 288 1C, DO2228 ppm 3.4E11NWC 288 1C, DO39ppm 6.4E12NWC 288 1C, DO39ppm 1.3E11NWC 288 1C, DO39ppm 1.0E11NWC 288 1C, DO39ppm 1.0E11
1.E-12
1.E-11
100101
C
K (MPam0.5)
Fig. 8. CGR of the 316LSS 0.5TCT specimen prepared from BWR core
shroud mock-up [14].HWC 288 1C, DO100150ppb, DH4050 ppb 1.5E12NWC 288 1C, DO39ppm 2.0E12
-
3.2. Irradiated materials: actually used BWR core shroud
and top guide
Microscopic views of surfaces of the SCC specimen areshown in Fig. 9. For HAZ area in the shroud, ditches ongrain boundaries were observed 23mm away from theweld fusion line, suggesting that the area was sensitized.For the base metal in the shroud and the top guide, noevidence of material degradation due to radiation wasfound, although some inclusions and/or precipitationsattributed to the heat treatment during the steel manu-facturing were observed.Table 4 shows the results from the tensile tests of
the irradiated components. The higher the uence causes,the higher the yield strength and UTS, and the lower theuniform elongation.Table 5 shows results of the SCC growth rate tests. The
specimens were pre-cracked by fatigue aiming at a/W 0.5in air and then in 288 1C water after water chemistrybecame stabilized (i.e., the autoclave outlet conductivitywas o0.2 uS/cm). Crack length was monitored andcorrected in the manner stated in the previous section.Crack length for the crack growth rate calculation was
ARTICLE IN PRESSH. Yamashita et al. / International Journal of Pr586estimated by dividing the intergranular crack area on thefracture surface by the specimens thickness, since weldingresidual stress was supposed to be released when specimenswere prepared from HAZ of actual core shrouds. Stressintensity factor, Kapplied, was calculated from load andspecimen geometry in all conditions.Fig. 10 shows the IASCC growth rate acquired under
NWC (32 ppm DO) in this study. All of the data are below
Fig. 9. Microscopic observation of the actually used core shroud [14].
Table 4
Tensile test results of the irradiated components (at 288 1C) [14]
Material Fluence
(n/m2)
Yield
strength
(MPa)
UTS
(MPa)
Uniform
elong. (%)
Strain to
failure (%)
SUS316 (T-1) 1.3E+25 630 737 12.7 20.9
1.3E+25 599 705 13.1 20.3
SUS316 (T-1) 5.4E+24 400 603 25.5 29.6
5.4E+24 404 601 22.1 28.8
SUS304 (S-2) 1.6E+24 321 517 21.5 27.6
1.6E+24 311 513 22.4 27.6SUS304 (S-3) 3.6E+24 477 559 13.5 19.6
3.6E+24 469 547 14.2 19.9the upper limit of SCC growth rate of unirradiatedsensitized 304SS in NWC, 9.2 1010m/s. Figs. 11 and12 show that the crack growth rates certainly decreaseunder HWC (2 ppb DO+2040 ppb DH). This factindicates that the mitigation by improving water chemistryis still effective even for SCC growth rates of irradiatedmaterials up to 1.2 1025 n/m2.
4. Discussion
4.1. Unirradiated material: H3 and H4 weld mock-up
Fig. 13 shows the hardness of the H3 weld mock-up andactually used core shroud of the H7a weld. Since theuence of H7a is as low as around 2.6 1019 n/m2, theradiation-hardening effect can be negligible. Their hardnessseems to be at a very similar level, around 200HV in HAZ,although some differences due to the kinds of weld and thedistance from the surfaces are seen. The mock-upssuccessfully simulate the properties of the actually usedcore shroud. Note that the hardness of the shroud isobviously low compared to that of PLR pipe welds whosehardness reaches as high as 270HV (Fig. 14, [3]), whichwould enhance the SCC growth rate.Consideration of (1) SCC growth rates with HAZ of the
core shroud mock-ups acquired in this study are envelopedby the disposition curve of JSME NA1-2002 standard forL-grade SS, and (2) SCC recently found in L-grade SSshroud initiated and propagated in HAZ, several mm awayfrom the fusion line [1,2], would naturally lead to theassumption that the above disposition curve is reasonablyapplicable even to the reactor core shroud only if theuence is low.
4.2. Irradiated materials: actually used BWR core shroud
and top guide
Fig. 15 shows the hardness increment of the irradiatedmaterials. In Figs. 15 and 16, external tube was the partof external tubes of dummy fuel assemblies irradiated in aJapanese boiling water reactor (this ux was about410 1016 n/m2s), and neutron source holders wereirradiated in a Japanese BWR for one and two cycles (theseux was about 5 1017 n/m2s). Although the data scatter, itcan be said that the irradiation effect tends to becomesmaller as the uence increases beyond a certain point,approximately 1.02.0 1025 n/m2. Further, if the materi-als are separated into the two group, (1) reasonable ux(shroud and top guide) and (2) high ux (others), it issuggested that the high ux promotes hardening for thesame uence. It can be thought that the density ofdislocation loops tends to increase with the increase inneutron ux at same uence [10]. Fig. 16 shows yieldstrength of the irradiated materials. Yield strengthincreased with neutron uence [10].
essure Vessels and Piping 85 (2008) 582592Since the difculty of SCC growth rate tests withsmall specimens was found through the past experiences,
-
ARTICLE IN PRESSPrTable 5
H. Yamashita et al. / International Journal ofspecimens as large as possible were made from the actuallyused core shroud and top guide, i.e., 1TCT/0.7TCTfrom the shroud and 9.1mm thickness 0.5T-CT from the
Condition and result of SCC growth rate test with the actually used core shro
Component
ID
SS Specimen
size
Specimen
no.
Fluencea
(n/m2)
Beffb
(mm)
syirrad288 1C
c
(MPa)
syeffectived
(MPa)
S-1 304 1T CT-1S 1.5E+24 22.7 318 239
S-1 304 CT-2S 1.6E+24 22.7 324 242
S-1 304
S-4 304 CT-3S 5.1E+24 22.7 454 307
S-4 304 CT-4S 4.5E+24 22.7 438 299
S-3 304 0.7T CT-5S 1.9E+24 15.9 341 250
S-3 304 CT-6S 2.4E+24 15.9 365 262
S-5 304 CT-7S 3.2E+24 15.9 396 278
S-5 304 CT-8S 4.4E+24 15.9 435 298
T-1 316 0.5T
9.1mmt
CT-1T 1.2E+25 8.1 583 371
T-1 316 CT-2T 1.2E+25 8.1 583 371
T-1 316 CT-3T 5.2E+24 8.1 457 308
T-1 316 CT-4T 5.2E+24 8.1 457 308
T-1 316 CT-5T 2.3E+24 8.1 360 260
T-1 316 CT-6T 2.3E+24 8.1 360 260
aFluence at the center of the specimen (ex. At the crack growth position, mbBeff (BBsg)0.5 [4].csyirrad at 288 1C was estimated using Fig. 16 (syirrad 2.867E5Fluencedsyeffective syunirrad.+(syirradsyunirrad.)/2, syunirrad. at 288 1C was assumeeKallowable syeffective*(Beff/2.5)0.5.fDeviation (KallowableKapplied)/Kallowable 100.gda/dt was based on the average crack length in constant load.hIn these test condition, the autoclave outlet conductivity was about 0.23
o0.2 uS/cm.)
1.E-12
1.E-11
1.E-10
1.E-09
1.E-08
100101K, MPam0.5
Crac
k G
row
th R
ate,
m/s
304 Base Metal (4.5-5.1E+24 n/m2)304 Base Metal (1.5-1.6E+24 n/m2)304 HAZ (4.4E+24 n/m2)304 HAZ (1.9-3.2E+24 n/m2)316 Base Metal (1.2E+25 n/m2)316 Base Metal (2.3-5.2E+24)
9.2x10-10m/s
32ppmDO
D.L. 2x10-12m/s
Fig. 10. CGR of irradiated 304SS and 316SS CT specimens prepared from
used BWR core shroud and top guide [14].ud and top guide [14]
Kallowablee
(MPam0.5)
Environment Kapplied(MPam0.5)
Deviationf
(%)
da/dtg
(m/s)
22.8 32 ppmDO 22.5 1.2 8.0E1223.1 32 ppmDO 10.5 54.5 o2.0E12
32 ppmDO 24.1 4.5 3.1E112 ppbDO+20ppbDH 24.2 4.9 9.1E12
29.3 32 ppmDO 11.7 60.0 1.0E102 ppbDO+20ppbDH 11.8 59.7 5.6E11
28.5 32 ppmDO 23.3 18.2 2.5E102 ppbDO+20ppbDH 23.5 17.5 2.2E12
20.0 32 ppmDO 12.0 39.9 o2.0E1220.9 32 ppmDO 12.0 42.6 o2.0E1222.2 32 ppmDO 13.5 39.2 o2.0E1223.7 32 ppmDO 12.4 47.7 5.6E1121.2 32 ppmDO 23.2 9.6 5.6E10h
20 ppbDO 24.0 13.3 6.3E102 ppbDO+40ppbDH 24.3 14.8 2.2E12
21.2 32 ppmDO 22.5 6.3 4.4E10h20 ppbDO 23.0 8.6 3.6E102 ppbDO+40ppbDH 23.2 9.6 2.8E11
essure Vessels and Piping 85 (2008) 582592 587top guide, so that SCC growth rate tests with high Kbecome valid. In addition, ux for the top guide,2.4 1016 n/m2s is reasonably close to that for the coreshroud, 12 1016 n/m2s. Consequently, considering thefact that those data were acquired using such largespecimens, it seems that the crack growth rates forirradiated materials shown in Table 5 are one of the mostreliable data to predict IASCC growth rate in the actualcore shroud.For all test of the irradiated materials, K validity was
evaluated using the following equations [4]. For irradiatedmaterials, use of the yield strength in Eq. (1) is non-conservative [4]:
Kallowable sy Beff=2:50:5, (1)
sy syunirrad: syirrad: syunirrad:=2. (2)Table 5 shows K validity and SCC growth rate data. As
shown, the K validity was mostly satised. There ispossibility that invalid specimens provided unreasonablyhigh SCC growth rate for the evaluation of core shroudwhere a small crack is assumed to propagate under elastic
17.6 32 ppmDO 13.5 23.2 o2.0E1217.6 32 ppmDO 12.6 28.3 o2.0E12
2 ppbDO+20ppbDH 12.8 27.2 o2.0E1214.8 32 ppmDO 10.5 29.2 o2.0E1214.8 32 ppmDO 10.5 29.2 o2.0E12
aximum uence of CT-4S and CT-8S is about 5E+24n/m2).
0.2914).
d to be 160MPa [4].
uS/cm. (In other tests conditions, the autoclave outlet conductivity was
-
ARTICLE IN PRESS
0.3mm
Fatiguepre-crack
IG
Fatiguepost-crack
Cons
Crac
k Le
ngth
, mm
24.5
24.3
24.1
23.9
23.7
23.5
Fig. 11. An example of the fracture surface (left) and crack growth trend (rig
shroud of 304SS (CT-3S) [14].
0.3mm
Fatigue pre-crack
IG
Fatigue post-crack
Crac
k Le
ngth
, mm
0.000Triang
0
12.8
12.6
12.4
12.2
12
11.8
Sp
Fig. 12. An example of the fracture surface (left) and crack growth trend (right
guide of 316SS (CT-1T) [14].
100
150
200
250
300
350
-2Distance from fusion line, mm
Mic
ro-V
icke
rs H
ardn
ess,
HV
K-3 H7a(Ring) at 1.5mm from inner surfaceK-3 H7a(Ring) at 4mm from inner surfaceH3 Mock-up(Ring) at 1mm from outer surfaceH3 Mock-up(Ring) at 12.5mm from outer surface
Base Metal
WeldMetal
K-3 H7a: 316L Fluence ~2.61019n/m2
0 2 4 6 8 10
Fig. 13. Hardness test result of the H3 weld mock up and actually used
core shroud of H7a weld [12].
H. Yamashita et al. / International Journal of Pr588Specimen : CT-3S, Fluence 5.1x1024n/m2, 1TCT
ht) of SCC growth rate test with a specimen made from the actually used0conthrrea
(1)
(2)
(3)
TK4in Fuelist
DO
4Hz Rle Wa
ecim
) of1.0E-10m/s
Time, h200 400 600 800 1000 1200
Kave=11.7MPa*m0.55.6E-11m/sKave=11.8MPa*m
0.532ppm DO 2ppbDO+20ppbDHConstant Loadtant Load0.0002Hz R0.7Triangle Wave
essure Vessels and Piping 85 (2008) 582592straint due to the surrounding material. The followingee facts, however, suggest that all of the tests were undersonable K condition [11].
Obvious strain hardening occurs beyond the yield pointas seen in the stressstrain curves of these componentsmaterials (Fig. 17).Difference between the allowable maximum K and theapplied K is small, i.e., (Kapplied/Kallowable1)/Kallowableis less than 15% (Table 5, deviation).Data of CT-1S and CT-2S (304SS base metal specimenswith low uence) in NWC are enveloped by the dis-position curve of JSME NA1-2002 standard forL-grade SS in NWC.
he crack growth rate data acquired in NWC under20MPam0.5 were picked up from the table and plottedig. 18. The crack growth rates increase with increasingnce. Actually, these data envelope all of the other dataed in the table.
Time, h
5.6E-10m/sKave=23.2MPa*m
0.5
6.3E-10m/sKave=24.0MPa*m
0.5
2.2E-12m/sKave=24.3MPa*m
0.5
DO
(ppm
, In
ret)
Constant Load0.7ve
100 200 300 400 500 6000.001
0.01
0.1
1
10
100
en : CT-1T, Fluence 1.2x1025n/m2, 0.5TCT
SCC growth rate test with a specimen made from the actually used top
-
ma600gui
ARTICLE IN PRESS
10
100
1000
1.0E+24
Har
dnes
s inc
rem
ent,
HV
Shroud (304)Top guide (316)Neutron source holder (316L)External Tube (304)
Fluence, n/m21.0E+25 1.0E+26
Fig. 15. Relationship between uence and hardness increment [7,10].
100
1000
1.0E+24
Fluence, n/m2
Yie
ld S
tress
at 2
88C,
MPa
Shroud (304)Top guide (316)External Tube (304)Neutron source holder (316L)
1.0E+25 1.0E+26
Fig. 16. Relationship between uence and yield strength at 288 1C [7,10].
Fig. 14. Hardness test results of PLR piping [3].
H. Yamashita et al. / International Journal of Prdifference in ux between the core shroud and the othermaterials, as seen in Fig. 22. It is necessary to consider thatthe material of the top guide is 316SS, but this guresuggests that the high ux promotes hardening for thesame uence.
4.3. Residual stress measurements of irradiating bead-on-
plate samples
In addition, high uence introduces the radiation-induced stress relaxation for the weld residual stress. Thiseffect is experimentally conrmed by irradiating bead-on-plate specimens in JMTR to about 1 1024, 5 1024,life.A2byC growth rate for the shroud even at the end of the plant
dditionally, the above assumption does not count thetheSCterial) for that uence is assumed to be aroundMPa, which is close to that of the actually used topde tested in this study. That leads to an assumption thatlatter half of 1010m/s order could reasonably encloseFig. 19 shows the hardness of a boat sample of thecore shroud (1F-4, H4 weld position, SUS304L,1.3 1025 n/m2). Since the nominal hardness of the un-irradiated material is 131HV, radiation hardening occurredobviously. As a result, no signicant difference was foundbetween the base metal and HAZ, adjacent to the weldfusion line.The above fact along with the following two additional
facts suggests that it is reasonable to estimate the SCCgrowth behavior in an actual shroud based on the SCCgrowth rate data with the base metal even if SCCpropagates in HAZ.
(1) Cr depletion does not occur due to the weld in L-gradeSS [6] and
(2) although the hardness of HAZ in the un-irradiated H3and H4 weld mock-ups increase up to around 200HV,such increment in hardness seems not to signicantlyaffect the crack growth rate (Fig. 8).
Fig. 20 shows almost linear relationships between thehardness and the yield strength [7], and Fig. 21 shows theyield strength and SCC growth rate. Data from un-irradiated wrought materials [4,8,9] are also plotted.The maximum crack growth rate in this study is around
78 1010m/s in high yield strength range. In addition,the sensitized wrought 304SS and non-sensitized wroughtSS show the similar trend in the high yield strength range.Therefore, it is assumed that work hardened non-sensitizedSS of which yield strength is above 600MPa could besubstituted for irradiated SS with similar yield strength.Incidentally, the maximum uence of H4 weld of the
core shroud for 1100MWe BWR plant reaches about1.5 1025 n/m2 after 60 years of operation. According toFig. 16, yield strength at 288 1C of 316L SS (shroud
essure Vessels and Piping 85 (2008) 582592 5891025 n/m2 followed by the residual stress measurementsthe neutron diffraction method. The specimens were
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ARTICLE IN PRESS
1.E-12
1.E-11
1.E-10
1.E-09
1.E-08
1.0E+24
Crac
k G
row
th R
ate,
m/s
304 Base Metal (Core Shroud)316 Base Metal (Top Guide)
NWC, K > 20MPam0.5
Fluence, n/m21.0E+25 1.0E+26
Fig. 18. Relationship between CGR and uence [14].
Fig. 17. Stress strain curve of the actually used core shroud (left) and top guide (right) [14].
0
100
200
300
400
-5Distance from fusion line, mm
at 3mm from inner surface
at 5mm from inner surface
WeldMetal Base Metal
Fluence ~1.3x1025n/m2
304L
HV of base metal refered from mill sheetM
icro
-Vic
kers
Har
dnes
s, H
V
0 5 10 15 20
Fig. 19. Hardness test results of core shroud (1F-4 H4) boat sample [12].
200
250
300
350
400
400Yield Stress at 288C, MPa
Ausenite SS (around 2E+25n/m2)
Vic
kers
-Har
dnes
s, H
V
500 600 700 800 900
Fig. 20. Relationship between hardness and yield strength in austenite SS [7].
1E-12
1E-11
1E-10
1E-9
1E-8
0Yield Stress at 288-300C, MPa
Crac
k G
row
th R
ate,
m/s
Sen 304SS (Andreseen (2003), Castao Marn (2003) Fig. 12L-SS WR (Andressen(2003), Shoji(2003) No or small martensite)347SS WR (Castao Marn (2003) Fig. 12)Shroud (304)Top guide (316)
YS ~600MPa 316L estimated to60 year used Shroud H4 (1100MW)
Unirradiated Materials K = 27.5 ~ 30 MPam0.5
Shoji (2003) K2.161 conversion @ 30 MPam0.5
200 400 600 800
Fig. 21. Relationship between yield strength and crack growth rate [4,8,9].
H. Yamashita et al. / International Journal of Pressure Vessels and Piping 85 (2008) 582592590
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ARTICLE IN PRESS, H
V Top Guide (316, Fluence2.1E+24n/m )Top Guide (316, Fluence4.7E+24n/m2)Top Guide (316, Fluence1.1E+25n/m2)
Pr0
50
100
150
1.0E+15
Vic
kers
har
dnes
s inc
rem
ent
External Tube (304, Fluence2.6E+24n/m2)External Tube (304, Fluence5.0E+24n/m2)External Tube (304, Fluence7.4E+24n/m2)External Tube (304, Fluence8.6E+24n/m2)
Flux (n/m2s)1.0E+16 1.0E+17200 Shroud (304, Fluence 2.3E+24n/m2)
Shroud (304, Fluence4.6E+24n/m2)Shroud (304, Fluence6.4E+24n/m2)
2
H. Yamashita et al. / International Journal ofmade of 304SS. Fig. 23 shows that the stress relaxation ofthe welding residual stress depends on the radiation dose.At 1.5 1025 n/m2 (about 2.1 dpa the maximum uence ofthe core shroud for 1100MWe BWR plant after 60 yearsoperation), the welding residual stress was about 60% ofone at un-irradiated conditions [13].K-value, which is a parameter of SCC growth rates, also
decreases by the weld residual stress relaxation by neutronradiation. So, this stress relaxation process should also beconsidered in assessing the SCC growth rates of BWR coreshrouds H4 HAZ.
5. Conclusion
The SCC growth rate tests and the fundamental propertytests were performed with the specimens made of un-
Fig. 22. Effect of ux on incremental hardness: data of which uences are
close are linked [10].
0
0.2
0.4
0.6
0.8
1
0Dose (dpa)
Bead on plate spacimens
Stre
ss re
laxa
tion
( /
0)
1 2 3 4
Fig. 23. Irradiation dose dependence of stress relaxation for peak stress
on mid-thickness of the bead-on-plate specimens (304SS). [13].irradiated H3 and H4 weld mock-ups, the actually usedcore shroud and the top guide. The results are as follows:Un-irradiated material: SCC growth rates of the un-
irradiated H3 and H4 weld HAZ were enveloped by thedisposition curve of JSME NA1-2002 standard, althoughthe hardness of HAZ in these areas increases up to around200HV. Such increment in hardness seems not tosignicantly affect the crack growth rate.Irradiated material: Hardness of the actually used core
shroud and top guide as well as the external tube increaseddue to the neutron irradiation. The increment of the formeris smaller than that of the latter.SCC growth rate tests with 1TCT specimens from the
core shroud BM, 0.7TCT specimens from the core shroudHAZ and 0.5TCT (with 9.1mm thickness) specimens fromthe top guide indicated that all of the rates are under theupper limit, 9.2 1010m/s of the disposition curve ofJSME NA1-2002 standard for sensitized 304 SS.Welding residual stress relaxation by neutron radiation:
The core shroud H4 HAZ of the 1100MWe BWR plantafter 60 years operation, the welding residual stressdecreases about 60% by neutron radiation.
Acknowledgments
This study was performed as a joint research programamong Japanese BWR utilities, Hitachi Ltd., ToshibaCorp. and Nippon Nuclear Fuel Development Co., Ltd.
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ARTICLE IN PRESSH. Yamashita et al. / International Journal of Pressure Vessels and Piping 85 (2008) 582592592
SCC growth behavior of BWR core shroud materialsIntroductionExperimentUn-irradiated material: H3 and H4 weld mock-upIrradiated materials: actually used BWR core shroud and top guide
ResultsUn-irradiated material: H3 and H4 weld mock-upIrradiated materials: actually used BWR core shroud and top guide
DiscussionUnirradiated material: H3 and H4 weld mock-upIrradiated materials: actually used BWR core shroud and top guideResidual stress measurements of irradiating bead-on-plate samples
ConclusionAcknowledgmentsReferences